ML022280270

Initial Submittal of Written Examination with NRC Comments - Form ES-401-7 for the Dresden Initial Examination - June 2002

ML022280270
Person / Time
Site:	Dresden
Issue date:	06/03/2002
From:	Otten M Exelon Generation Co, Exelon Nuclear
To:	Office of Nuclear Reactor Regulation
References
50-237/02-301, 50-249/02-301, ES-401-7 NUREG-1021, Rev 8
Download: ML022280270 (161)

Text

{{#Wiki_filter:Fo-- ES-401-7 Wrtten Examination Quality Checklist Facihity: -/v/ Date of Exam. /-T/C .I Exam Le.,-R "*al Item Descnption aI: c'

1.

Questions and answers technically accurate and applicable to facility x -

2.

a. NRC K/As referenced for all questions

b. Facility learning objectives referenced as available

3.

RO/SRO overlap is no more than 75 percent, and SRO questions are appropriate per Section D.2.d of ES-401

4.

Question selection and duplication from the last two NRC licensing exams ,:.£..-. appears consistent with a systematic sampling process

5.

Question duplication from the license screening/audit exam was controlled as indicated below (check the item that applies) and appears appropriate: athe audit exam was systematically and randomly developed; or the audit exam was completed before the license exam was started; or the examinations were developed independently; or /1' _'ihe licensee certifies that there is no duplication; or -_ other (explain)

6.

Bank use meets limits (no more than 75 Bank Modified New percent from the bank at least 10'percent new, and the rest modified); enter the actual question 3 distribution at right

7.

Between 50 and 60 percent of the questions on Memory CIA the exam (including 10 new questions) are written at the comprehension/analysis level; enter the actual question distribution at right Z/ 2.---" ,4,

8.

References/handouts provided do not give away answers 0

9.

Question content conforms with specific K/A statements in the previously approved examination outline and is appropriate for the Tier to which they are .4 assigned; deviations are justified

10.

Question psychometric quality and format meet ES, Appendix B, guidelines

11.

The exam contains 100, one-point, multiple choice items; the total is correct and agrees with value on cover sheet Printed Name/ Si nature Date

a. Author y

b. Facility Reviewer ()

4i..

c. NRC Chief Examiner (M) 7L 1/ A&*

d. NRC Regional Supervisor Note:

The facility reviewer's initials/signature are not applicable for NRC-developed'examinations

1. Independent NRC reviewer initial items in Column "c;" chief examiner concurrence require:.

NUREG-1021, Revision 8, Supplement 1 42of46 Exam Submittal ES-401

ES-401 W'itten Examination Form ES-401-7 Quality Checkist Facility: Date of Exam. rA 7-Exam Level ; R RQ Initial Item Des:'potion a b° c 1 Questions and answers technically a=-urate and applicable to facility

2.

a. NRC K/As referenced for all quest zc-s

b. Facility learning objectives referenced as available

/tt .L L2-2.

3.

RO/SRO overlap is no more than 75 percent, and SRO questions are appropriate ,441) ./ per Section D.2.d of ES--401 V

4.

Question selection and duplication froi the last two NRC licensing exams .j-t4 appears consistent with a systematic sampling process

5.

Question duplication from the license screening/audit exam was controlled as inflicated below (check the item that aDolies) and appears appropriate: Z the audit exam was systematically aid randomly developed; or the audit exam was completed befo-e the license exam was started; or ,the examinations were developed irdependently; or licensee certifies that there is no duplication; or other (explain)

6.

Bank use meets limits (no more than 75 Bank Modified New percent from the bank at least 10 percent new, and the rest modified); enter the actual question 7 2.. -'7 4c) distribution at right 3 7

3.

7 F

7.

Between 50 and 60 percent of the questions on Memory C/A the exam (including 10 new questions) are written at the comprehensiort/analysis ievel;...."- f K enter the actual question distribution at right

8.

References/handouts provided do not give away answers ,,4fL) 4

9.

Question content conforms with specific K/A statements in the previously approved examination outline and is appropriate for the Tier to which they are assigned; deviations are justified

10.

Question psychometric quality and format meet ES, Appendix B, guidelines 4

11.

The exam contains 100, one-point, mubple choice items; the total is correct and As4} ý'-, I agrees with value on cover sheet Printed Name / Si/pnature Date a, Author Z_**

b. Facility Reviewer (

c. NRC Chief Examiner(#)

'zr4__

d. NRC Regional Supervisor 4iZt

__c Note: The facility reviewer's initials/signature are not applicable for NRC-developed examinations.

1. Independent NRC reviewer initial items in Column "c;" chief examiner concurrence required.

NUREG-1021, Revision 8, Supplement 1 42 of 46 Exam Submittal

E "-401 BWR SIM Examination Outline F-,cilitv: Dresden Exam Date: 05/27/2002 Printed: 02/072002 Form ES-401 -1 Exam Level: SRO K/A Category Points TI-Group Point Total K I K2 K3 K4 K5 K6 A l A2 A3 A4 G 1.4 5 3 5 5 4 26 ergency 2 3 2 3 3 3 3 17 .,.-normal Plant Tier "z-olutions Totals 7 7 6 8 8 7 43 1 2 2 2 2 2 2 2 2 2 2 23

2.

2 1 1 2 1 1 2 1 0 1 0 3 13 Plant _. stems 3 0 0 0 0 1 0 0 1 0 0 2 4 Tier Totals 3 3 4 3 4 4 3 3 3 2 40 Cat I Cat 2 Cat 3 Cat 4 -. Generic Knowledge And Abilities 4 5 4 4 17 ?/ te:

1. Attempt to distribute topics among all K/A Categories; select at least one topic from every K/A category within each tier.

2. Actual point totals must match those specified in the table.

3. Select topics from many systems; avoid selecting more than two or three I1KA topics from a given system unless they relate to plant-specific priorities.

4. Systems/evolutions within each group are identified on the associated outline.

5. The shaded areas are not applicable to the category tier.

Exam Sub- -tal I

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( fijic Iiicl 0,( _U) and Abnormal Plant Evolutions - Tier 1/ Group 1 K2 K3 Al KA Topic X AA I0" - M ES - 401 4J 4j E col E x w I 1-aci.llll 3: 1 )1 ckclJI I 'II [LI'.'. 0 -11()/1 _00 -1

Facility: Dresden BWR SRO Examination Outline Printed: 02/07/2002 ES - 401 Emergency and Abnormal Plant Evolutions - Tier I / Group 1 I KIIK2IK3IAIA2 G KATonic 4j 44 E D o3 E x Xi

Facility: Dresden BWR SRO Examination Outline Printed: 02/07/2002 ES - 401 K/A Category Totals: 4 5 3 5 5 4 Group Point Total: 26 t 44 4*J E E x )<

BWR SRO Examination Outline Dresden Facility: ES - 401 E/APE # 295001 Printed: 02/07/2002 295002 295002 295004 295005 295005 295008 295019 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 2 E/APE Name / Safety Function KI K2 K3 Al A2 G KA Topic Partial or Complete Loss of Forced Core Flow X 2.2.25 - Knowledge of bases in technical specifications Circulation / 1 for limiting conditions for operations and safety limits. Loss of Main Condenser Vacuum / 3 X 2.4.49 - Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. Loss of Main Condenser Vacuum / 3 X AK2.04 - Reactor/turbine pressure regulating system Partial or Complete Loss of D.C. Power /6 X AK'3.02 - Ground isolation/fault determination Main Turbine Generator Trip / 3 X AA 1. 04 - Main generator controls Main Turbine Generator Trip / 3 X AK1.02 - -ýCore thermal limit considerations High Reactor Water Level / 2 X AK3.04 - Reactor feed pump trip: Plant-Specific Partial or Complete Loss of Instrument Air / 8 X AA2.02 - Status of safety-related instrument air system loads (see AK2.I-AK2.19) Partial or Complete Loss of Instrument Air /8 X AK2.17 - High pressure coolant injection: Plant-Specific Loss of CRD Pumps / I X AA2.01 - Accumulator pressure HigLh Drywell Temperature / 5 X EK1.02 - Equipment environmental qualification High Suppression Pool Water Level / 5 X EA2.03 - Drywell/containment water level 3.6 3.1 3.5 Form ES-401-1 Imp. Points 3.7 1 4.0 1 3.3

3. 3 2.8 3.6 3.5 3.7 29u0 19 295022 29(5028 29 S02 9 4.E E

M to E III 2.7 1

BWR SRO Examination Outline Facility: Dresden ES - 401 Fency and Abnormal Plant Evolutions - Tier 1 / Group 2 KI K2 K3 Al A2 G KATopic X 2.4.49 - Ability to perform without refercncC to procedures those actions that require immediate operation of system components and controls. EA 1.03 - Secondary containment ventilation Printed: 02/07/2002 K/A Category Totals: 3 2 3 3 3 3 Group Point Total: 17 4~J E Co E X x

BWR SRO Examination Outline Facility: Dresden ES - 401 Printed: 02/07/2002 4gJ E D X w

BWR SRO Examination Outline Printed: 02/07/2002 Facility: Dresden System / Evolution Name Automatic Depressurization System / 3 218000 Automatic Depressurization System / Primary Containment System and Auxiliaries / 5 Primary Containment Isolation System/Nuclear Steam Supply Shut-Off/ 5 Primary Containment Isolationl System/Nuclear Steam Supply Sliut-Off/ 5 RI1IRI PCI: Containment Spray System Mode/ 5 R*eIc-WeCr: CoLetaiventr Spray System Mode / 5 Reactor Water Level Control System/ KI X K2 K3 X X ES - 401 Sys/Ev # 218000 223001 223002 223002 .`6001 226001 239002 Plant Systems - Tier 2 / Group I 6 Al A2 A3 A4 G IKATopic X X K4 K5 X I_ 4 4 K5.02 - Water hammer T r i-i- I X A2.01 - Loss of any number of main steam flo inputs I I InputsI X A,I 06 - DP/S6Wl 'lantpc"clIcmcHI c -l selector Switch: P'lant-Specific K3.01 - Restoration of reactor water level after a break that does not depressurize the reactor when required K4.01 - Prevent inadvertent initiatior of ADS logic A3.02 - Vacuum breaker/relief valve operation 2.4.4 - Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. K I. 19 - Component cooling water systems K2.02 - Pumps Form ES-401-1 Imp. IPoints Imp. Points I ___________________ X 4.4* 3.9 1 3.4 4.3 2.9 1 2.9Q* 2.7 L _____ E .0 E (0 X x K6 I I I 3.4 I

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BWR SRO Examination Outline Printed: 02/07/2002 Facility: Dresden ES - 401 Plant Systems - Tier 2 /Group 1 Form ES-401-1 Svs/Ev # System / Evolution Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 261000 Standby Gas Treatment System / 9 X A2.04 - High train moisture content 2.7 1 262001 A.C. Electrical Distribution / 6 X K2.01 - Off-site sources of power . 3.6 1 264000 Emergency Generators (Diesel/Jet) / 6 X A 1.09 - Maintaining minimum load on 3.1 1 emergency generator (to prevent reverse power) K/A Category Totals: 2 2 2 2 2 2 2 2 2 2 3 Group Point Total: 23 E4~ E X W

BWR SRO Examination Outline Printed: 02/07/2002 Facility: Dresden ES - 401 Plant Systems - Tier 2 / Group 2 Form ES-401-1 Sys/Ev # System / Evolution Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic Imp. Points 201001 Control Rod Drive Hydraulic System X K2.05 - Alternate rod insertion valve solenoids: 4.5* Plant-Specific I< WR) ( I'li,pcclc ) / /IP-Spec(Not-BWR6) 202001 Rccircu lnton System / I X K6.02 - Component cooling water systems 3.2 1 204000 Reactor Water Cleanup System / 2 X 2.1.14 - Knowledge of system status criteria 3.3 which require the notification of plant personnel. 205000 Slhutdowvn Cooling System (RHR X K4.03 - Low reactor water level: Plant-Specific 3.8 Shutdown Cooling Mode) / 4 215002 Rod Block Monitor System / 7 X K1.01 - APRM: BWR-3, 4, 5 3.0 219000 RHR!LPCI: Torus/Suppression Pool X A 1.02 - System flow 3.5 Cooling Mode / 5 230000 RHR/LPCl: Torus/Suppression Pool X K6.05 - Suppression pool 3.4 Spray Mode 5 4 4000 Fuel Handling Fquipment /8 X 2.1.2 - Knowledge of operator responsibilities 4.0 during all modes of plant operation. 2,:1000 Main Turbine Generator and X K3.02-Reactor pressure 4.0 Auxiliary Systems,' 4 E to E co X hi

BWR SRO Examination Outline Printed: 02/07/2002 Facility: Dresden K/A Category Totals: 1 1 2 1 1 2 1 0 1 0 3 Group Point Total: 13 4-J 4-4 E D

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X w

BWR SRO Examination Outline Printed: 02/07/2002 Facility: Dresden ES - 401 K/A Category Totals: 0 0 0 0 1 0 0 1 0 0 2 Group Point Total: 4 4-J E 5<

(,*encric l e Iowh 'te a1ld \\ Alilitites ()hitline (Fier 3) BWR SRO Examination Outline Facility: Dresden Printed: 02/07/2002 Form ES-401-5 %, eneric C.ategory KA KA Topic Imp. Points Conduct of Operations 2.1.13 Knowledge of facility requirements for controlling vital / controlled access. 2.9 1 2.1.11 Knowledge of less than one hour technical specification action statements for systems. 3.8 1 2.1.22 Ability to determine Mode of Operation. 3.3 1 2.1.8 Ability to coordinate personnel activities outside the control room. 3.6 1 Category Total: 4 Equipment Control 2.2.3 (multi-unit) Knowledge of the design, procedural, and operational differences between 3.3 1 units. 2.2.8 Knowledge of the process for determining if the proposed change, test, or experiment 3.3 1 involves an unreviewed safety question. 2.2.26 Knowledge of refueling administrative requirements. 3.7 1 2.2.2 Ability to manipulate the console controls as required to operate the facility between 3.5 1 shutdown and designated power levels. 2.2.34 Knowledge of the process for determining the internal and external effects on core 3.2* 1 reactivity. Category Total: 5 Radiation Control 2.3.9 Knowledge of the process for performing a containment purge. 3.4 1 2.3.6 Knowledge of the requirements for reviewing and approving release permits. 3.1 1 2.3.2 Knowledge of facility ALARA program. 2.9 I 2.3.4 Knowledge of radiation exposure limits and contamination control, including permissible 3.1 1 levels in excess of those authorized. Category Total: 4 4-J 4.J E E X W I

"teneric Nnowlecige and Abilities Outline (Tier 3) BWR SRO Examination Outline Facility: Dresden Printed: 02/07/2002 Form ES-401-5 Generic Category Emergency Plan 2.4.7 KA KA Topic 2.4.32 fKnowledge of operator response to loss of all annunciators. 2.4.45 Knowledge of event based EOP mitigation strategies. Imp. Points 3.8 1 Knowledge of local auxiliary operator tasks during emergency operations including 3.5 1 system geography and system implications. Ability to prioritize and interpret the significance of each annunciator or alarm. 3.6 1 Category iotal: 4 Generic Total: 17 4-, 4-, E E M LU 2 I

ES-401 F!cilityj Dresclen I'R R) RO Examination Outline Exam Date: 05/27/2002 Printed: 02'07 2002 Form S.-401-2 Exam Level: RO K/A Category Points Tier Group K1 K2 K3 K4 K5 K6 A! \\2 A3 A4 G Point Total I 4 3 3 2 1 0 13

1.

Emergency 2 4 4 3 4 3 1 19 Abnormal Plant 3 0 0 1 2 0 1 4 Evolutions Totals 8 7 7 8 4 2 36 Tier 1 2 3 3 3 2 2 2 3 3 3 2 28

2.

2 1 2 2 2 3 2 2 2 2 1 0 19 Plant Systems 3 0 0 0 0 0 0 0 1 1 0 2 4 Tier Totals 3 5 5 5 5 4 4 6 6 4 4 51 Cat I Cat 2 Cat 3 Cat 4

3. Generic Knowledge And Abilities 3

3 4 3 13 Note: I. Attempt to distribute topics among all K/A Categories; select at least one topic from every K/A category within each tier.

2. Actual point totals must match those specified in the table.

3. Select topics from many systems; avoid selecting more than two or three K/A topics from a given system unless they relate to plant-specific priorities.

4. Systems/evolutions within each group are. identified on the associated outline.

5. The shaded areas are not applicable to the category tier.

E, '- Subm i tta I I

,den E/APE # 295005 295005 295014 295014 295015 295024 295025 295031 295031 295037 295037 500000 500000 BWR R" tamination Outline Printed: 02/( Facility: ES - 401 E/APE Name / Safety Function Main Turbine Generator Trip / 3 Main Turbine Generator Trip / 3 Inadvertent Reactivity Addition / I Inadvertent Reactivity Addition / 5 Incomplete SCRAM / I Highi Drywell Pressure /5 High Reactor Pressure /3 Reactor Low Water Level / 2 Reactor Low Water Level / 2 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown / I SCRAM Cond ition Present and Reactor Power Above APRM.Downscale or Unknown I High Containment Hydrogen Concentration/ 5 High C~ontainment Hydroge~n C onc.entration / 5 K/A Category Totals: 4 3 3 2 1 0 Group Point Total: 13 Emergency and Abnormal Plant Evolutions - Tier 1 / Group 1 Ki K2 K3 Al A2 G KATopic X AA 1.04 - Main generator controls X AK1.02 - tCore thermal limit considerations X AK2.01 - RPS X AK3.02 - Control rod blocks X AK1.04 - Reactor pressure: Plant-Specific X EK1.01 - Drywell integrity: Plant-Specific X EK3.04 - Isolation condenser initiation: Plant-Specific X EK2.16 - Reactor water level control X EA2.01 - Reactor water level X EKI1.02 - Reactor water level effects on reactor power X EA1.04-SBLC X EK2.09 - Drywell nitrogen purge system X EK3.01 - Initiation of containment atmosphere control system Form ES-401-2 Imp. Points 2.7 3.2 1 .3.9 1 3.74 1 3.8 4.1 4.5* 4.1" 1 4.6*

4. 1*

4.5~ 3.0 2.9 I I

m x 3 Facility: SES - 401 E/APE # rt Ft a295002 E/APE Name / Safety Function Loss of Main Condenser Vacuum / 3 Partial or Complete Loss of A.C. Power / 6 Partial or Complete Loss of A.C. Power / 6 Partial or Complete Loss of D.C. Power / 6 295008 High Reactor Water Level / 2 High Suppression Pool Temperature / 5 t Control Room. Abandonment / 7 High Off-Site Release Rate / 9 I l Sgupre I ssi-on PclWa.tc Leve 9 Partial or Complete Loss of Component Coolinzg Water / 8 Partial or Complete Loss of Instrument Air /8 Inadvertent Containment Isolation/5. Hilgh DryweUl Temperature / 5 Ifigh Suppression Pool Water Level /5 BWR RC tamination Outline Printed: 02/( 295003 ,den 295003 295004 295013 295016 295017 29 8I / 295018 295019 295020 295028 295029 I I Emergency and Abnormal Plant Evolutions - Tier 1 / Group 2 K1 K2 K3 Al A2 G KATopic X AK2.04 - Reactor/turbine pressure regulating system X AK2.03 - A.C. electrical distribution system X AA 1.03 - Systems necessary to assure safe plant shutdown X AK3.02 - Ground isolation/fault determination X AK3.04 - Reactor feed pump trip: Plant-Specific X AA2.01 - Suppression pool temperature X AA 1.04 - A.C. electrical distribution X AK2.04 - Plant ventilation systems X AK3.03 - iHIciplcnientation of sitc cicrgeicy plan X 2.1.14 - Knowledge of system status criteria which require the notification of plant personnel. X AK2.17 - High pressure coolant injection: Plant-Specific X AA2.01 - Drywell/containnient pressure X EKI.02 - Equipment environmental qualification X EA2.02 - Reactor pressure Form ES-401-2 Imp. Points 3.2 1 3.7 1 4.4* 1 .9 3.3 3.8 3.1 3.3 2.5 2.7 1 3.6 2.9 3.5

BWR RC -amination Outline Printed: 02/C and Abnormal Plant Evolutions - Tier 1 / Group 2 K2 K3 Al A2 G KATopic X FAl 0 1 - Aren rlPA t K/A Category Totals: 4 4 3 4 3 1 Group Point Total: 19 x 21 Facility: 3 ,den

m x a) 3 Facility: SES - 401 E/APE # rt rt a 295023 Printed: 02/C Fornm ES-401-2 Imp. Points Refueling Accidents / 8 295023 Refueling Accidents / 8 X High Secondary Containment Area Temperature / 5 Secondary Containment High Differential Pressure / 5 x I i I I I Ix AA1.03 - Fuel handling equipment 2.1.14 - Knowledge of system status criteria which require the notification of plant personnel. EAI.03 - Secondary containment ventilation EK3.02 - Secondary containment ventilation response K/A Category Totals: 0 0 1 2 0 1 Group Point Total: 4 ,den Ki K2~K~TAniA -t I-UI -r E/APE Name / Safety Function BWR RC amination Outline Emergency and Abnormal Plant Evolutions - Tier 1 / Group 3 295032 295035 2.5 1 3.7 3.3 K1 K2IK31A1 A2 X X

x Facility: _,'esden c-ES - 401 Sys/Ev # System / S20100N 1

• .Tr BWR RO r "mination Outline Printed:

0: Plant Systems - Tier 2 / Group 1 K6 Al A2 A3 A4 C KATopic K2.05 - Alternate rod insertion valve solenoids: Plant-Specific "002

BWR RO F -nination Outline Facility: Dresden Sys/Ev # System/Evolution Name K1 K2 K3 K4 K5 K6 Al A2 A3 A4 G KA Topic 215004 Source Range Monitor (SRM) System X A4.04 - SRM drive control switches /7 Average Power Range Monitor/Local Power Range Monitor System / 7 Average Power Range Monitor/Local Power Range Monitor System / 7 Nuclear Boiler Instrumentation / 7 T Nuclear Boiler Instrumentation / 7 Automatic Dcpressurization System / 3 X X X X I I I_ I L Automatic Depressurization System / 3 Primary Containment System and Auxiliaries / 5 Primary Containment Isolation System/Nuclear Steam Supply Shut-Off / 5 X X f-[---I I I I I.- I l I_ I__ X X A3.07 - RPS status K5.06 - Assignment of LPRM's to specific APRM channels K6.01 - A.C. electrical distribution K4.01 - Reading of nuclear boiler parameters outside the control room K3.01 - Restoration of reactor water level after a break that does not depressurize the reactor when required K4.01 - Prevent inadvertent initiatior of ADS logic A3.02 - Vacuum breaker/relief valve operation K I. 19 - Component cooling water systems Form ES-401-2 Imp. Points 3.2 1 3.8 3.1 3.6 4.4* 3.7 3.4 2.7 ~ ~. L - I.______I ES - 401 Printed: 02 902 215005 215005 216000 216000 218000 218000 223001 223002 i i i i I I I I I I_ I 1 1 I

BWR RO F rmination Outline ES - 401 Sys/Ev # System / Evolution Name 241000 Reactor/Turbine Pressure Regulating System / 3 Reactor Feedwater System / 2 _5 9002 Reactor Water I.cvcl Control System 259002 Reactor Water Level Control System 261000 261000 2 Standby Gas Treatment System / 9 Standby Gas Treatment System / 9 KI I K2 K4 K5 i i i 264000 Emergency Generators (Diesel/Jet) /6 264000 Emergency Generators (Diesel/Jet) / 6 K/A Category Totals: 2 3 3 3 2 Plant Systems - Tier 2 / Group 1 K6 Al A2 A3 A4 G KA Topic X 2.4.4 - Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. X 2.1.2 - Knowledge of operator responsibilities during all modes of plant operation. X A2.01 - Loss of any ntunber of main steam 110 inputs X A4.06 - DP/Single/three element control selector switch: Plant-Specific X A2.04 - High train moisture content X A2.13 - High secondary containment ventilation exhaust radiation X A 1.09 - Maintaining minimum load on emergency generator (to prevent reverse power) X A3.06 - Cooling water system operation 2 2 3 3 3 2 Group Point Form ES-401-2 T_---1 hnp. 4.0 Points I 370 I

3.

1 315 2.5 3.0

3.1 Total

28 Facility: Liresden 259001 Printed: 02 002 3

BWR RO F mination Outline Printed: 02, 002 Facility: Dresden ES - 401 Sys/Ev # System / Evolution Name KI K2 K3 K4 IK5 Plant Sy~om er roup C,.. stm Ti / G 2_______________________________________ K6 IA1 A2 A3 A4 !G KATnnic Form ES-401-2 _ _Imp. lPOints 201006 Rod Worth Minimizer System x A3.05 - Latched group indication: 3.0 1 (RWM) (Plant Specific) / 7 P-Spec(Not-BWR6) 202001 Recirculation System / I x K6.02 - Component cooling water sys tems t-Specific 205000 Shutdown Cooling System (RHR X K4.03 - Low reactor water level: Plan Shutdown Cooling Mode) (4 205000 215002 219000 219000 226001 226001 230000 Shutdown Cooling System (RHR Shutdown Cooling Mode) / 4 Rod Block Monitor System / 7 RIR!LPCI: Torus/Suppression Pool Cooling Mode 5 RHR/LPCI: Torus/Suppression Pool Cooling Mode/ 5 R.HR/LPCI: Containment Spray System Mode*/ 5 RHR/LPCI: Containment Spray System Mode 15 RHR/LPCI: Torus/Suppression Pool Spray Mode / 5 x x x Main and Reheat Steam Systemn / 3 x x x x A4.11 - Heat exchanger cooling flow K1.01 - APRM: BWR-3, 4, 5 A 1.02 - System flow A2.05 - A.C. electrical failures K2.02 - Pumps K5.02 - Water hammer K6.05 - Suppression Pool K5.06 - Air operated MSIV's 3.1 1 3.8 1 3.2 1 2.9 I 3.5 3.3 1 2).9* 1 2.6 1 3.3 I 2.8 t 239001 X

BWR RO F mination Outline Printed: 02/ ý02 System / Evolution Name A4 I G I KA Tonic Main Turbine Generator and Auxiliary Systems / 4 K1 IK2 IK3 1K4 X K5 Plant Svstems - Tier 2 / Zrnu,n I~ K6 Al. A2 IA3 K3.02 - Reactor pressure i t1t11-I1f1I i-f-{---1 I I____ Reactor Condensate System / 2 T A.C. Electrical Distribution / 6 A.C. Electrical Distribution / 6 Uninterruptable Power Supply (A.C./D.C.) /6 X X X IX 271000 Offgas System /9 X K5.07 - Radioac 290003 Control Room HVAC /9 -]X-j A3.01 - Initiatioi A2.12 - Loss of equipment component cooling water systems K2.01 - Off-site sources of power A 1.01 - Effect on instrumentation and controls of switching power supplies K3.17 - Process monitoring: Plant-Specific 2.9 1 tive decay n/reconfiguration Vel to other, *ili sv;NIt:III-; K4.02 - Cross-ov Group lPoint Total: 19 Facility: Dresden ES - 401 Sys/Ev # 245000 256000 262001 262001 262002 Form ES-401-2 Imp. Points 3.9 1 3.1 D.3 3.1 .)00000 Instrument Air System ([AS) / 8 L K/A.. a.. r

1.

2 2 2 3 2 2 2 2 K/A Category Totals: 1 2 2 2 3 2 2 2 2 2.7 3.3 I I I I I A4 G KATonic 1 0

M x G) 3 Facility: , esden C, ES-401 B Sys/Ev # System / rt BWR RO r mination Outline Printed: 02 002 K/A Category Totals: 0 0 0 0 0 0 0 1 1 0 2 Group Point Total: 4 I

Generic Knowledge and Abilities Outline (Tier 3) BWR RO Examination Outline Printed: 02/07/206 Form ES-401-5 KA KA Topic Conduct of Operations 2.1.22 Ability to determine Mode of Operation. 2.8 1 2.1.8 Ability to coordinate personnel activities outside the control room. 3.8 1 2.1.32 Ability to explain and apply system limits and precautions. 3.4 1 Category Total: 3 Equipment Control 2.2.26 Knowledge of refueling administrative requirements. 2.5 1 2.2.2 Ability to manipulate the console controls as required to operate the facility between 4.0 1 shutdown and designated power levels. 2.2.34 Knowledge of the process for determining the internal and external effects on core 2.8 reactivity. Emergency Plan 2.3.2 2.3.4 2.3.1 2.3.10 I _______________________ 2.4.7 2.4.35 2.4.45 J I Category Total: 3 Knowledge of facility ALARA program. 2.5 T -1 Knowledge of radiation exposure limits and contamination control, including permissible 2.5 1 levels in excess of those authorized. Knowledge of 10 CFR 20 and related facility radiation control requirements. 2.6 1 Ability to perform procedures to reduce excessive levels of radiation and guard against 2.9 1 personnel exposure. Category Total: 4 Knowledge of event based BOP mitigation strategies. 3.1 Knowledge of local auxiliary operator tasks during emergency operations including system geography and system implications. Ability to prioritize and interpret the significance of each annunciator or alarm. 3.3 Category Total: 3 Generic Total: 13 Facility: Dresden Generic Category RtuiLitOH Co. 0ntrol I Emergency Plan Imp. Points

ES-401 Written Examination Form ES-401-9 Review Worksheet The first 30 questions were used for the sample. All 125 questions were reviewed for accuracy and K/A match.

1.

2.

3. Psychometric Flaws

4. Job Content Flaws

5. Other

6.

7.

Q# LOK LOD 1 t 1 (F/H) (1-5) Stem Cues T/F ICred. Partial Job-Minutia

1. /

Back-Q= SRO U/E/S Explanation Focus J Dist". Link units ward K/AOny 1 H 2 S 2 F 2 S 3 H 2.5 E Put stem in past tense. Move The from each distractor to the stem. 4 H 2 E Put stem in past tense. 5 H 2 S 6 F 2 E Add ? symbol at the end of the question. 7 H 2 E Change the stem to read, "The cooldown rate is controlled by throttling the..." and delete repeated material in each distractor. Move "at" from the end of the stem to correct location within stem. Facility changed distractors to eliminate multiple correct answers. 9 H 2.5 S 10 H 2.5 S Instructions Refer to Section D of ES-401 and Appendix B for additional information regarding each of the following concepts.]

1.

Enter the level of knowledge (LOK) of each question as either (F)undamental or (H)igher cognitive level.

2.

Enter the level of difficulty (LOD) of each question using a 1 - 5 (easy - difficult) rating scale (questions in the 2 - 4 range are acceptable).

3.

Check the appropriate box if a psychometric flaw is identified: The stem lacks sufficient focus to elicit the correct answer (e.g., unclear intent, more information is needed, or too much needless information). The stem or distractors contain cues (i.e., clues, specific determiners, phrasing, length, etc). The answer choices are a collection of unrelated true/false statements. More than one distractor is not credible. One or more distractors is (are) partially correct (e.g., if the applicant can make unstated assumptions that are not contradicted by stem).

4.

Check the appropriate box if ajob content error is identified: The question is not linked to the job requirements (i.e., the question has a valid K/A but, as written, is not operational in content). [h,, l "i,.J i III,, I t r<, lit...

th, 1 i,+,.-;III "f k 11, m l-II,It lh +'1:1 i,: h1 *, I.< ill,* l,I 1,i I l, 1 z, " I "tl 11, l

• l,

I-:* I I< 11i, l< " 1, I il II,: II' i ,,l r,, l~d it 1 h, I 1, 1, l WI ,w l hI,t r, r *, I 1=, " i ..,, itl+*i.u ),l Il,, - illl, i, It,,,,i iý: ii. I..... 1 4

• y,..

it ",.- J ;: i 1-. 1,,. il:ý d< tJ. I + ' ` 1 i,, t -11 I,-,,, ill 1. +, li, -r i,. d, r H. p II hU L k; ZAItu tu l I UL*IltllL: Uý I Vtjl:ýU Ictlm. +l,,.I pl~l -ll;ilull I l 1l),1 klltt 1(' 1111, 11Al) I ~l t~l~ltl1 lli[lI[:,.

5.

Check questions that are sampled for conformance with the approved K/A and those that are designated SRO-only (K/A and license level mismatches are unacceptable).

6.

Based on the reviewer's judgment, is the question as written (U)nacceptable (requiring repair or replacement), in need of (E)ditorial enhancement, or (S)atisfactory?

7.

At a minimum, explain any "U" ratings (e.g., how the Appendix B psychometric attributes are not being met).

ES-401 2 Form ES-401-9

1.

2.

3. Psychometric Flaws

4. Job Content Flaws

5. Other

6.

7.

Q# LOK LOD (F/H) (1-5) Stem Cues T/F I Cred. Partial Job-Minutia

1. /

Back-Q= SRO U/E/S Explanation Focus J Dist. Link units ward K/A Onl 11 F 2.0 E Change stem to past tense. 12 H 2.0 S 13 H 2.5 S 14 H 2.5 E Put the stem in past tense. 15 F 2.0 S 16 F 2.0 S K/A asks power supply to RPS MGs, question gives power supply. 17 F 2.0 US Station says Higher, 0 is memory. Licensee demonstrated that question matches K/A, and is higher skill question. 18 H 2.5 19 H 2.5 S 20 F 2.0 S 21 H 2.5 S 22 F 2.0 S 23 H 2.0 S 24 F 2.0 S 25 H 3.0 S Station added 'MSIVs are closed" to current plant conditions. 26 H 2.5 S 27 F 2.0 S Doesn't satisfy the "predict" portion of the K/A. Does satisfy procedure 28 H 2.5 X k)S Predicts because of knowing what equipment is left. 29 H 2.0 S

ES-401 3 Form ES-401-9

1.

2.

3. Psychometric Flaws

4. Job Content Flaws

5. Other

6.

7.

Q# LOK LOD (F/H) 1(1"5) Stem CuesIT/F Cred. Partial Job-Minutia

1. /

Back-Q= SRO U/E/S Explanation j J__ Focs Jj Dist. LinkllunitsJ ward JK/A ONI 317 F 2.0 SIZID__II WI I_ __IIsI 324H 2.0 rIzII IWEDYI____s_ 33] F 2.0 ] I I XI I I I I I S Candistractor a. be construed as a correct answer? a. was changed 34]H]30 ]__ I ZIZI S I I IZIIsl 35 F 2.0K SI__7]_I I I IZI__S 36 H 2.0 II II I I E Put the stem in past tense. 37 H 2.0 I I I i I I II I I ZII Put the stem in past tense. 38 H 2.0 1 I I I I I II S l 40 1 2.0 I I IZ I Z II I I IF 20 EI put the symbol "+in front of all the inches (column 2). L411 H 1I3.0J I Z I 1 SEW I [42 HI 2.51 SZIZ 1IIIZ I________ 43 1 H 3.0 1 II I IIII S I 44 H 1 2.5 1 II I II I I I I S

• 145.1 HI 2.

I I S Is 46 1H 2.01 LIZWII IIIIZZI I__ I471F 2.01 [IIZ IZI__III IZ I 148 1 1 1 H I S I 249 1 2.0 1 11 I I IZ I I I I sI..I

ES-401 4 Form ES-401-9

1.

2.

3. Psychometric Flaws

4. Job Content Flaws

5. Other

6.

7.

Q# LOK LODTI 1 1 (F/H) (1-5) Stem Cues T/F Cred. Partial Job-Minutia

1. /

Back-Q= SRO U/E/S Explanation Focus Dist. Link lunitsjward K/A SOnly

• o !_1_o l1II I1I..I.I.I.I I Is I _ _ _ _ _ _ _ _ _ _ __

I 1I1I olI I I Z Z II I II I I III I_ IL.I_ _ Is 1_______________________ _1_1_ I I I I_ __ __ I I __ _1 1 1 1 ] _ I ___Is l _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ is3I.I1I5I I__ I 1 I__ I ___I I I__ 1 I__ 1 I u I Nothigher. Only requires identification of memorized items. Agree 1Z1 1 *11 11i...Z Z ___L....I II __II I_ I __I _[___Is l

• 1I. 1z5I I

III I I I__ Ii I~ Ii i ___ I _________s l_______________ I61 I1I 1 I IL I _ I _ IIi __ 1 1 1 ] [ I __ Is l

2. r I I I1lI I 1 Ihl Qdesn't match K/A. K/A addresses ability, 0 addresses knowledge.

57 H l.0de 572 H_2.5 x U__ Traveling s speled travelng. Reworked answers to require ability. 585III 12"0I _ I I III L I I 2.5_S I* 1' 1*5 ___1 ] I I I I... I..I I 11 I Is l 6'I1" 1 0 1I ___l...I...I I_ I I__ I I I__ 1 1 1 1 s I S___________________I____ 6F 1 I o _ I I I I sI I I Isple E JRework the first line into two statements. 60 H 2.5 1 I 11 I S l 6 1 1I.0 1 I I I I I I I I 0I I Is 61 IZ11.0 1 1II L I I I__ II I11 [_ I Is l 66II 1I.oIZ I I I I Ii I*..j1I I_ I [ I Is l 6___.5 I I_ 111 I I__ I1 1I11 I Is l 68 1 IH I IE [ework Question deals with acceptable values, K/Awith highpooltemps Have 68 Hus 1 I Questo know acceptable values as well as non-acceptaple, Good as is.

ES-401 5 Form ES-401-9

1.

2.

3. Psychometric Flaws

4. Job Content Flaws

5. Other

6.

7.

Q # L O K L O D I [x1l1n1ti o n (F/H) (1-5) Stem Cues T/F Cred. Partial Job-Minutia

1. / IBack-Q=

SRO U/E/S Explanation Focus Dist. Link units ward K/A Only I 0 1 II t ] I _ I _ I_ IJ I _ iThis question has two correct answers. b. needs to be changed. Agree F69 H _12.0 S that b. is not correct because of initial conditions. I7 I1" !I 1I I__ 1 __ I I_ T I __I I 1iI~ I __ Is _ 1_ 1 F I~I I I]I__ I I_ ]I.... I__ ]I1 I Is l I~73.I*.*IZ IZIi.... I .....I I....I __ ]((1I1lu I Not higher. Question is memory level. Agreed 7I F 2.51 IIZ I IZ I IZI I Isl 75 1 11 I 1 1 I 1((T1 high off-site release rate - vs -ventilation systems? After explanation, 75F [.5_ IS agree with question as written. 76_FI[o [III))[ I 2.0 I I 77 F 2.o0 I II I I I I I I I Is l 78T H 1 iI T I us Not higher. Question is memory level. Special case, fails opposite of 78 H 3.0 _S expected. 79 F 2.5 [ II I I I I 1 IZI I I S l V*o 1 [. 51

• 1 II

_ __I I I _ _ I I_ [ I 1 _ I _ _ _ _ ___s l_ 80 H 2 [ 82 H 2.0 S 83H __0 1 1III I 2.0 [ ! I I ]s 841 .I [I I ZI H J I [ I I 1 I2I U iNot higher. Question is memory level. Agreed. I85 H 2.0 S[I I 1ii 1__ i__] 861 H 12.5 I [ I I I _ I I 1 U I Maybehigher. Facilityagrees-questionishigher 87f H 1 2.0 1I II I I II I S Is

ES-401 6 Form ES-401-9

2.

3. Psychometric Flaws

4. Job Content Flaws

5. Other 1 6.

7.

Q# (F/H) (-5) Stem Cues T/F Cred. Partial Job-Minutia

1. /

Back-Q= SROI U/E/S Explanation Ste Focus T' Dist. Link units ward K/A Only F88~r IH1251 1 1 1 1 11 T TI f I S Mayb two correct answers. HPCI exhaust goes to the torus. IS Explained. Answer c. is not correct. _8 H1 211 II I I I I H I. 1I I E [Put the stem in past tense. o90 1j~J J__ I I 25 S I I I ishouldn't the alarm be 903-4? Agreed ,IF *1 __ 1Z __III __11 _]IJ_ 1] sI __31F I o I1 11I __ I___ I I_ I I__ 1 Z 1 I I s l _9__F

2.

1 III I II I0I I I Isl 196 1 F 1 o 2.0 1I I I I I S l 197 1 I I IHI III 20 I J e 1Put the stem in past tense. Re-word where necessary. 98 [ F 2.0 U I I _ 1 I I IZ I I I U why is a. wrong? a. was correct, changed a. F 200 II I FI I III I 2.011S I~Loo71o 2. _IIII .I ___ ]1I1 I _ I~ I__ I IsI 11011 F 1 2.0 1I I IIIIIII I 11S 1102 [ H 1 2. 1 _ IZ.1 _ 1 1il I Z ~L s I________________________ Io1

• 103 1

FI 1 2.0 1 1 1 S Is 1104F1 0 1 IZII 1_1_ I Z IISI[I I Isl 106 LH i2.0 1 1 1 I aNot higher. Question is memory level. Minimally higher. Comparison I.......[ ........ ( I ________L__________] with a setpoint is required to get the correct answer.

ES-401 7 Form ES-401-9

1.

2.

3. Psychometric Flaws

4. Job Content Flaws

5. Other

6.

7.

Q# LOK LOD (F/H) (1-5) Stem Cues T/F Cred. Partial Job. Minutia

1. / IBack-Q=

SRO U/E/S Explanation Focus Dist. Link units ward K/A On y [07 F 2.0 x r I [1 I answer doesn't match distracts Changed to SRO question. corrected 118 2 x U Hanswer to correct wording. H109 1F lZl II I I I I I I I I SIs [*o F 1 2.0o IZ IZ I I _ I 1 1 __ I] _ _ 1 1 I__ I I s l F o 1 1 I [T I I

c.

and d. are arguably the same response. Clarified "on shift operator" [ F 20 S so c. is not correct. 11111 F 2.0 Z ZI II S__V 1 1__ 1112J F 2.0 S__ 1 __ [ 1131 F 2.0 IZIiIZIZIII I I I I lu Question is a memory level question. Agreed [114 F 1 2.01 1 1 T 1I I aMay bemudsUl le corrrect answers (RO/SRO required by license?) Qis ok 1 1 15 1_ FJL _1 2.0_ 11 1_ aS st n s 5 N s required for the calculations, not R O /S R C. 151 F I II I I I I I I I Hs [5 1 117 1 F 12.5 IZ Z Z Z 1__ 1.... 1_ 1_ 1..... I 1_ _ 111 F 2.5 I I I I I I I ZI I S Is F11 I9I lI I t T I I[ l U change "a administrative" to "an administrative" This is a memory q. 1120 2.5 U Agreed, not higher. 1121 I I I I I I I I I I I 12,1 1HI3.0 IZ IZ J I ZI I I I__ 11 __ I- ]___ I Is________________________ 1=12 IH 2.5 IZ I I IZ I I SI I ! I s J typofor the answer. its not'e.' 112IFIoIZ ZI I III I FI I 21S _ _4 1 2I.01 I I I_ I __ I I I I__ II ~ Is l 11251 F 1 2.0 I I I I i I I I s Tmake sure the other answers are incorrect. vernfied -others are wrong.

rouudu of Rejectea KJAs Fcwrn FR..Afltin Tier / Group BOTH 1/1 BOTH 2/3 SRO 2/1 SRO 2/2 RO 2/3 r 4 NUREG-1021, Revision 8, Supplement 1 ES 401 Forrn Fq-*A n i-idn Randomly Se ected K/A 295007 3.0 290002 3.06 261000 ..30 234000 .03 Tier / Group BOTH 1/1 BOTH 2/3 SRO 2/1 SRO 2/2_ RO 2/3 Randomly Selected K/A 295007 AK3.01 290002 K3.06 261000 2.4.30 234000 A2.03 Reason for Rejection Similar to Question #86 K/A 295025 EK3.04 replaced with 295031 EA2.01 Replaced due to inapplicability to Dresden replaced with 288000 2.1.33 Similar to Question #71 K/A 295015 2.4.30 replaced with 215005 2.2.22 Replaced based on applicability comments from exam validators and facility representative, replaced with 234000 2.1.2

Dresden 2002 IIT NRC FExam QU Exaim System 1K,'A RO SRO I BOTI f 201001 K2.05 4.5 4.5 Control Rod Drive Hydraulic System Objective: 2 12L-S2-06 Knowledge of electrical power supplies to the Alternate rod insertion valve solenoids mAnt-Specific following: The following events occurred in sequence from rated conditions on Unit 2: 125 VDC control power to Bus 22 was lost. A full reactor scram signal was received. Control rods were inserted by actuation of ARI Which ONE of the following describes the expected response to the ARI initiation? The ARI valves in: A. division 1 energized. B. division 2 energized. C. both divisions energized. D. both divisions DE-energized. ANSWER:

a.

division I energized. Explanation: ARI is an "energize to actuate" feature. With a loss of control power from Bus 22, the Divisio.-i 2 ARI valves cannot energize. Only the Division 1 ARI valves will energize on the manual initiation.

Reference:

SDM 212002 Question Pedigree: BANK 21200S0541 Cog level: High R, I 03/28/02 1 of I Exam Suomittal

Jeln 2002 ILT NRC' lxam

)06 KA.0 A 3.05 Rod Worth Minimizer System (RWN1

ant Specific)

Ability to monitor automatic operatio: ROD WORTH MINIMIZER SYSTE'. including: IObjective. 201 L-S6 03, 20106LP0-05 the Latched group indication: P-Spec(Not-BWR6) RWM) After taking the shift, it is noticed that - Rod Worth Minimizer for control rod F-6 is displaying a rod position of 28 in GREEN. This indica:-: -hat the rod... A. has an insert error. B. is in an unknown position. C. is in the current latched step. D. has an alternate limit assigned. 03/28/02 I of I Exam Submittal Q# 2 Rxam BOTlI RO I3.0 SRO Explanation: ANSWER

c.

is in the current latched step. Explanation: Insert error is Magenta, Unknown posi,:. 1 is Red ??, and Alternate limit is Yellow

Reference:

SDM 201006 Question Pedigree: Modified 20106S0181 Cog level: Memory Rev 2 Ob ective: 201L-$6-03, 20106LP005

Drcsden 2002 lT NRC F!xa\\ Q',l Fxam System K, A RO SRO 3 ,301'1 1 202001 K6.02 3.1 3.2 Recirculation System 1 Ob jective: 202L-S1-12b Knowledgc of the effect that a loss or Component cooling water systems malfunction of the following will have on the RECIRCULATION SYSTEM: Unit 2 is operating at rated conditions when the following events occur: The disc inside 2-3713 B-500 (2A RECIRC PMP OUTER SEAL CLR RBCCW INLET VLV) separates from the stem. Alarm "2A RECIRC PP SEAL CLG WTR FLOW LO" comes in. All other RBCCW parameters are normal. If no operator actions are taken the... A. 2A Recirc pump seals and bearings could be damaged within one minute. B. TCVs on the RBCCW system will open to lower the RBCCW temperature. 1C. 2A Recirc pump seats will operate normally as long as CRD flow is maintained. D. RWCU system could isolate since cooling is lost to the non-regenerative beat exchanger. Explanation: ANSWER

a.

2A Recirc pump seals and bearings could be damaged within one minute. Explanation: B is wrong since there are no indications that RBCCW temperature is rising. C is incorrect because with cooling flow lost damage could occur with or with out CRD flow. (Common misconception is that with CRD flow lost damage occurs in a short amount of time.) D is wrong since the RWCUs are on a different loop of the RBCCW system.

Reference:

DOA 3700-01 and DAN 902(3)-4 G-3 Question Pedigree: NEW Cog level: High Rev 2 03/28/02 1 of I Exam Subn ttal

l)resden 2002 1 LT NRC 'xam QiE xam System K/A RO SRO 4 BOTIl 202002 K4.05 3.1 3.4 Recirculation Flow Control System Ojcie2021ý-$2-05ý Knowledtge of RECIRCULATION IFLOW Limiting recirculation pumnp -speed mismatch: Plat CONTROL SYSTEM design feature(s) and/or I Specific interlocks which provide for the following:I The recirculating pumps are operating in the Master Manual mode at 75% speed when the Speed Contro the 3A pump saturates high. Because of this, the 3A... Explanation: ANSWER:

a.

pump increases to 85% speed and stops. Explanation: The pump mismatch circuitry will-stop any further pump speed increases once the mismatch exceeds 10%.

Reference:

SDM 202002 Question Pedigree: BANK 20202S0 151 Cog level: High Rev I I of I 03/28/02 Exam SubmittaI

Dresden 2002 IT NRC I\\xam BOTI I System 202002 Recirculation Flow Control System Knowledge of the operational implications of the F following concepts as they apply to RECIRCULATION FLOW CONTROL SYSTEM: The amount of oil inside the impeller casing of the M, the reactor is at maximum thermal power. 03/28/02 I of I Exam Submittal K A K.01 RO 2.8 SRO 2.8

l)rcoSICn 2002 I1,T NRC lxam Q# X11am System K,' A RO SRO 6 <RO( 204000

2. I 14 2.5 3.3 Reactor Water C. :up System Objective: 20400K1100I Conduct of'Oper...

ns Knowledge of system status criteria which require tile notification of plant personnel. For which of the owing events or conditions must the Plant Manager be notified according to OP-AA-l06 101 "Significant E. -nt Reporting" A. Failure of SPES B. An automatic.-elation of the RWCU system. C. Condensate ci Jttistry sample results indicate Action Level 1 parameters. D. A contractor is Found to be contaminated when going through the monitor at the gatehouse. Explanation: ANSWER

b. An automatic iso 3tion of the RWCU system.

Explanation: Per OP-AA-106-10' the failure of SPDS does not require notification, Action level 2 or above is required for notification, and the contractor found to be contaminated does not require notification.

Reference:

OP-AA-I06-101 Question Pedigree: New Cog level: Memory

Rev I 03/28/02 1 of I Exam Submittal

Dbresden 2002 II L NRC \\xain Q"" Fxam System K/A RO SR(J 7 RO 205000 A4.I I 3.2 3.2 Shutdown Cooling System (RIIR Shutdown Cooling Mode) Objective: 20500LK008 Ability to manually operate and/or monitor in Heat exchanuer cooling flow tthe control room: The following conditions exist on Unit 2: Unit 2 is shutdown. Reactor coolant temperature is 2000F. The Shutdown Cooling system is in service. Alternate Decay Heat Removal is NOT being used. How is the cooldown rate maintained? A. Throttling the RBCCW outlet from the heat exchanger with the SDC pump suction valve throttled. B. Throttling the SDC pump suction valve with the RBCCW outlet valve from the heat exchange open to the max position. C. Throttling the SDC pump discharge valve with tihe RBCCW outlet valve from the heat exchanger open to the max position. D. Throttling the RBCCW outlet from the SDC heat exchanger with the SDC pump discharge valve open to the max position. Explanation: ANSWER:

d. Throttling the RBCCW outlet from the SDC heat exchanger with the SDC pump discharge valve open to the max position.

Explanation: A is wrong because throttling is done with the SDC pump discharge valve. B and C are wrong because the RBCCW valve is full open above 212°F.The word max position are used in the distracters instead of open because the SDC pump discharge valves are only allowed to be open to approximately 60% due to limitations on the pumps.

Reference:

SDM 205000 Question Pedigree: Modified 20500S0011 Cog level: High Rev I 03/28/02 1 of I Exam Submittal

DrIesden 2002 II. T NRC b'xam QExam System K/A RO -0 8 BOTIFl 205000 K4.03 3.8 Shutdown Cooling System (RVIR Shutdown Cooling Mode) Objective: 205L-Sl-0-5 Knowledge of SHUTDOWN COOLING Low reactor water level: Plant-Specific SYSTEM/MODE design feature(s) and/or interlocks which provide for the following: Which of the following is the LOWEST reactor water level that the Shutdown Cooling syster tH normally operate at on Unit 3 A. +13 inches B. +5 inches C. -0 inches D). -8 inches Explanation: ANSWER:

a.

+13 inches. Explanation: The isolation for SDC occurs at +8 inches.

Reference:

SDM 205000 Question Pedigree: New Cog level: Memory Re-. 2 03/28/02 1 of I Exam Submittal

Dresden _ C) 11, 1NRC' Ia:m Sy steil) 206000 K/A A 1.06 lfigh Pressure Coolant Injection System Ability to predict and/or monitor changes in parameters associated with operating the HIG V PRESSURE COOLANT INJECTION SYSTE'.' (HIPCI) controls including: Syste Ob iective: 2061-SI -08; 206001,K002; 20600LK004 im flow: BWPR-2, 3, 4 A scram occurred on Unit 2 and the following -Inditions exist: 0 0 S 0 S 0 Feedwater is NOT available. Reactor level: -10" and dropping slow0. HPCI running in the pressure control n Ae. HPCI discharge pressure: 1100 psig co? -:ant. Reactor pressure: 850 psig and rising s:,ly. The Isolation Condenser is isolated due - a tube leak. Shortly after the Unit Supervisor directs the NS, to raise level with HPCI, the following occurs: S 0 0 HPCI flow increased rapidly. Reactor level quickly rose to +55". Annunciator 902-3 A9, HPCI TURB TRIPPED, alarmed. Which of the following describes the cause of t-ese conditions? A. The NSO throttled open the 2-2301-10 (tes: return valve) prior to injecting with HPCI. B. The HPCI Flow Controller output failed lo,, (zero output) while HPCI was injecting into the vessel. C. The MSC failed to the HSS. Turbine speed increased rapidly resulting in increased flow to the reactor. D. The NSO failed to reduce HPCI discharge r-essure to below reactor pressure prior to opening the 2-2301-8 (HPCI injection) valve. IxAp iai1a ANSWE d.

LIUII,

• R:

The NSO failed to reduce HPCI discharge pressure to below reactor pressure prior to opening the 2 2301-8 (HPCI injection) valve. EXPLANATION: The MSC is normally on the HSS when the mac.ine is running and speed is being controlled by the MGU. The MSC failing to the HSS will have NO effect. The NSO is suppose to open the 2-2301-10 valve to reduce HPCI pressure prior to injecting with HPCI. If the HPC[ flow controller fails low, the turbine will drive to it's slowest speed and pump discharge pressure will stabilizz between 375 and 600 psig. Of the alternatives listed, only the operator error )f failing to reduce HPCI discharge pressure before opening the "8" valve would cause the indications provided.

Reference:

DOP 2300-03 Qeto eire Question Pedigree: BANK 20600S0341 Cog level: Higl 03/28/02 Rev 2 I of I Exam Submittal 9 FBxar POTII I RO SRO 3.7

Dresden 2002 IL " NRCE' Ex vm )f Exam System KMA RO SRO '0 BOTIH 206000 IK6.Q0 3.5 3.5 Aigh Pressure Coolant Injection Syste ObJective: 206L-S 1-12 wsnowled-e of the effect that a loss orthentest ine. Talfunction of the following will have on the

• IGH PRESSURE COOLANT INJECTION SYSTEM (HlPCI):

Gjiven tihe following conditions; The plant is on line at 80% power

• A HPCI operability surveillance was in prolgress.
• HPCI was pumping 5200 gpmn to the CST through the test line.

'rile NSO shut the 2301-6 valve, HPCI CST Suction valve

• Annunciator 902-3 A-11, "fHPCI BOOST PP SUCT PRESS LO" has just alarmed.

NWhich of the following describes the response of the HPCR syste? .k. The HPCI turbine will trip on low booster pump suction pressure. B. The Flow Controller will decrease turbine speed until flow is zero. C. HPCI will continue to operate but pump flow will eventually drop to zero. D. The Flow Controller will increase turbine speed until it trips on overspeed. Explance: ANSWER:

a.

The HPCI turbine will trip on low booster pump suction pressure. EXPLANATION: The Booster Pump low suction pressure trip is bypassed with an initiation signal present. Because an initiation silgnal is NOT present, HPCI will trip; 'a' is correct and Vc is incorrect. There is no automatic scale back of

peed based on suction pressure; Vb is incorrect. The turbine speed control system CANNOT raise turbine 7speed to the overspeed setpoint. The turbine governor prevents this from occurring; Vd is incorrect.

Re ference: SDM 206000, DAN 902(3)-A I1I Question Pedligyree. BANK 20600S0292 Cog level: High Rev 2 C -,28/02 1 of I Exam Subm -- al

D)resden 2002 ILT NR" lxam Q4 Exam System K A RO SRO II RO 207000 K3.02 3.8 4.0 Isolation (Emergency) Condenser Objective: 07I,-5-12 Knowledge of the effect that a loss or Reactor water level (EPG's address the isolation condenser malfunction of the ISOLATION as a water source): BWR-2, 3 (EMERGENCY) CONDENSER will have on following: Unit 3 is at rated conditions when the following occurs: Loss of all high pressure feed. Isolation Condenser is being used for pressure control. Reactor water level is -65 inches and steady. A tube leak then develops in the Isolation condenser. This would result in.... A. Reactor water level lowering. B. A potential Group 4 isolation. C. Isolation Condenser shell side level lowering. D. Increased makeup flow to the Isolation Condenser. Explanation: ANSWER:

a. Reactor water level lowering.

The Iso condenser is at lower pressure than the reactor and water inventory would flow from the vessel to the iso condenser causing RPV level to lower.

Reference:

Question Pedigree: DOA 1300-1 and DAN 902(3)-3 H-2 New Cog level: Memory Rev 2 03/28/02 1 of I Exam Submittal

Dresden 2002 IILT NRC F\\. Q# [xam System K \\ RO SR0 12 BO1T1 209001 K I. 3.7 3.s Low Pressure Core Spray System hI 0.- v,e: 209[L-S 1-03 Knowledge of the physical connections and/or Emeii-rgecy iv 2 o r cause-effect relationships between LOW PRESSURE CORE SPRAY SYSTEM and the following: The following conditions exist: A complete loss ofoffsite power to BOTH units SA DBA LOCA on Unit 3 With NO operator action, what is the power supply to the 3A Core 5 -IV pump? A. U2 EDG B. U3 EDG C. U2/3 EDG D. U2 SBO DG 03/28/02 Exam Submittal I of I

[)resden 2002 I1, NRC F'x\\m F, -- I System 209001 Exam Submittal Q11 13 1K3.03 RO 2.9 S RO 3.0 Low Pressure Core System ObJective: 209L-SI-12 Knowledge of the e":- thcat a loss or Emergency generators malfunction of the I PRESSURE CORE SPRAY SYSTEM %, uave on following: While at rated condi: -s the following events occur: 01:28, A-- st due to a switching error. 01: 0, 2B-I st due to an electrical fire. 01:44, DBA I -CA occurs on Unit 2. Which of the follow\\:- .ill occur? A. U-2 EDG starts B. U-2/3 EDG stare5 C. CAM System B -:--..-ts D. SBO Feed at Buý _" trips Explanation: ANSWER:

b.

U-2/3 EDG c-ts Explanation Core Spray initiation :,gic is what sends a start signal to the EDGs on a DBA LOCA. 2B-I is the power supply to Division 2 of the C: re Spray logic. The Core Spray logic is what sends signals to other components in the plant to perform diffe- --t functions during a DBA LOCA. All of the components listed could be started by other means, howeve: cie loss of Core Spray logic prevents the Unit 2 EDG from starting, the CAM system B from starting, and pre. rnts the SBO Main Feed to Bus 24 from receiving the trip signal on a DBA LOCA on Unit 2

Reference:

SDM 209 .1 Question Pedigree: BANK 20901 S0262 Cog level: High Rev I 03/28/02 I ot I

Dresden 2002 II, NMC f,xam BXain SRO System 211000 KiA 2.4.6 RO 3.1 SRO 4.0 Standby Liquid Control System Objective: 2950211K040 Emergency Procedures and Plan Knowledge symptom based EOP mitigation strategies. During an Anticipated Transient Without a Scram on Unit 3, operators have lowered level to -140 inches. When the SBLC tank level has decreased to 39%, operators commence raising reactor water level per DEOP 400-5, Failure to Scram. While increasing reactor water level plant conditions are as follows: -operators are increasing reactor water level with Reactor Feed Pumps 0 -reaCLor pressure 5s 92u psig anu constant -reactor power is steadily increasing on IRMs Select the reason that reactor power increased when reactor water level was raised and what action needs to be taken. REASON A. Insufficient boron has been injected into the core to maintain shutdown conditions in reactor. ACTION Terminate and prevent all injection with the exception of boron and CRD. B. Insufficient boron has been injected into the core Continue to raise level to 8 inches and hold level to maintain shutdown conditions in reactor, between 8 and 48 inches. C. The water injected into the vessel flushed some Terminate and prevent all injection with the exception of the boron from the core area. of boron and CRD. D. The water injected into the vessel flushed some Continue to raise level to 8 inches and hold level of the boron from the core area. between 8 and 48 inches. Explanation: ANSWER:

a. Insufficient boron has been injected into the core to maintain shutdown conditions in reactor, and Terminate and prevent all injection with the exception of boron and CRD.

This condition is discussed in the BWROG EPG/SAG. Normally as level is increased the boron that was in the lower plenum now enters the core area causing power to decrease after a few seconds of increased flow. If power continues to increase it is because not enough boron has been injected for the condition the reactor is in currently. If these conditions exist DEOP 400-5 directs that all injections be terminated and prevented with the exception of boron and CRD.

Reference:

295L-S8 Qisii eire Question Pedigree: Bank 29502S 1051 Cog level: High Rev 2 I 03/28/02 I of I Exam Submittal 14

Dresdcn 2002 IL f NRC Exam Exam RO System 211000 Exam Submittal 15 K/A K2.02 RO 3.1 SR 03/28/02 I of I

Dresdc- " '2 HlT NRC Exam Q/1 16 Exam RO

Systen, 21200(

Reactor Protection System Ability to manually operate and/or monitor ir. the control room: 03/28/02 Exam Submittal K/A A4.06 4.2 SRO 4.1 r I

Q# 17 Exa in RO Dresden 2002 IL, TNRC Fxam System K/A 212000 K2.01 RO 13.2 SRO

3.3 Explanation

ANSVE R:

b.

A half scram will occur on RPS B logic channel as a result of the breaker trip. Explanation: MCC 28-2 provides power to RPS MG set A, which powers RPS Bus B, which powers RPS trip system B. RPS, being a deenergize to actuate system, then trips placing the unit in a half scram condition.

Reference:

DOP 500-3 and SDM 262011 Question Pedigree: Bank 21200S0331 Cog level: High Rev I 03/28/02 1 of I Exam Submittal

Dresden 2002 ILT NRC Exam System 215001 K/A A 3.03 T raversing In-Core Probe Ability to monitor automatic operations of the TRAVERSING IN-CORE PROBE including: Valve op I Objective: 215L-SI-5.a eration: Not-BWR I Automatic TIP traces are in progress on Unit 2 when a transient occurs resulting in the following conditions: RPV water level is +5 inches and rising. Drywell pressure is 1.5 psig and steady. Concerning the TIP system you would verify... A. the shear valve fires, isolating the TIP tube. B. TIP withdrawal to In-Shield position and Ball valve closure. C. the Group II Isolation status light on the TIP drawer is illuminated. D. the Shear AND Squib Valve Monitor lights are illuminated after 5 minutes. 03/28/02 I of I Exam Submittal Qt 18 Exam RO RO SRO

2.6 Explanation

ANSWER

b.

TIP withdrawal to In-Shield position and Ball valve closure. The Group II isolation signal (RPV level < +8 inches) would cause any TIP detector NOT in its shield to shift to manual reverse and withdraw into its shield chamber. Then the Ball valve would automatically close. Verifying these actions is a requirement of DAN 902(3)-5, E-5.

Reference:

Question Pedigree: SDM 215001 and DAN 902-5, E-5 MODIFIED 215 01SO0171 Cog level: High Rev 2 Valve or

Dresden 2002 II.TNRC Exam Q// Exam System K/A R( SRO 19 BOTH 215002 K 1.01 1.; 3.0 Rod Block Monitor System Objective: 2151 ---)5 and 06 Knowledge of the physical connections and/or APRM: BWR-3, 4, 5 cause-effect relationships between ROD BLOCK MONITOR SYSTEM and the following: Given the following conditions: Rod H-8 is selected. Reactor power is 40%. APRM Channel 3 fails "Downscale". APRM Channel 3 has NOT been bypassed. Due to this, the Rod Block Monitor (RBM) Channel 7... A. is NOT affected. B. is automatically bypassed. C. generates a rod withdrawal block. D. shifts to the alternate reference APRM. Explanation: ANSWER:

b.

is automatically bypassed. EXPLANATION: APRM 3 is the reference APRM for RBM Channel 7. There is no auto swap to th- 'Jternate APRM. When it fails downscale, RBM 7 thinks power is <30%. Power < 30% causes automatic b,-:ass of RBM Channel 7.

Reference:

Question Pedigree: SDM 215002 BANK 21502S0053 Cog level: High Rev 1 03/28/02 1 of I Exam Submittal

Dresden 2002 II.T NRC 1 amni Exam BOTH System 215004 Explanation: ANSWER:

c. momentarily depressed co-:inually held The "drive in" push button is a mraintaining contact and the "drive out" is a contact that must be continually held Reterence: SDM215004 Cog level: Memory Question Pedigree:

Modified question # 23 of last years NRC exam Rev I I of I Exam Submittal Ql; 20 KI/A A4.04 RO 3.2 SRO 3.2 I 03/28/02

Dresden 2002 II.T NRC Fxam Q# 21 Examn BOTH System 215005 Average Power Range Monitor/Local Power P Systern Ability to monitor automatic operations of the APRM/LPRM including: 03/28/02 I of I Exam SBbmittal K/A A3.07 RO 3.8 SRO 3.8

Dresden 2002 ILT NRC Exam Q4 Exam System K/A RO SRO 22 RO 215005 K5.06 2.5 2.6 Average Power Range Monitor/Local Power Range Monitor Objective: 2 15L-S5-03.b., 5.b. and 5.c. SystemI Knowledge of the operational implications of the Assignmient of LPRM's to specific APRM channels following concepts as they apply to APRM/LPRM: At least () LRsoncanl1,,3nd

2) on channels 4, 5, 6 of the LPRM's assigned to one APRM must be operable or an INOP trip occurs.

1 2 A. 20 21 B. 21 20 C. 10 11 D. 11 10 Explanation: ANSWER:

d.

11, 10

Reference:

SDM 215005 Question Pedigree: BANK 21505S0081 Cog level: Memory Rev 2 03/28/02 1 of I Exam Submittal

Dresden 2002 IL[ < Exam Q# Exam System K RO SRO 23 RO 216000 K-3.6 3.6 Nuclear Boiler Instrumentation Objective: 29501 LP083 Knowledge of NUCLEAR BOILER Readin:- ý nuclear boiler parameters outside the control INSTRUMENTATION design feature(s) and/or room interlocks which provide for the following: Given the following conditions: Unit 2 is at 80% reactor power when smoke begins. owing from the control room ventilation ducts. The Unit Supervisor directs a control room evacuat': Both Unit NSO's take all preparatory actions per DS..? 100-CR "Control Room Evacuation". As the Unit 2 NSO is leaving the control room, he n --- es RPV water level on the wide range indicator reading -68 inches. In order to continue monitoring RPV water level based on hI- -ast observation, the Unit 2 NSO should report to the: A. 5 or 6 racks in the Reactor Building. B. 7 or 8 racks in the Reactor Building. C. ATWS cabinets 2202-70A/B in the AEER. D. Analog Trip System racks 2202-73A/B in turbine buildi.- Explanation: ANSWER:

b.

7 or 8 racks in Reactor Building. EXPLANATION: The 7/8 racks provide local indication when level is below --6 inches. Above that level the NSO could go to the 5 or 6 rack.

Reference:

SDM 216000 and DSSP 100-CR Qu-,stion Pedigree: BANK 21600S0071 Cog level: High Rev 1 03/28/02 I of I Exam Submittal

Dresden 2002 I.,T NRC Exam Q# Nxam System K/ A RO SRO 24 30TH 216000 lK6.01 3.1 3.3 Nuclear Boiler 1-P :umentation Objective: 216L-S 1-06 Knowledge of ti:. fect that a loss or A.C. electrical distribution malfunction oft'-, 'bllowing will have on the NUCLEAR BO_-LR INSTRUMENTATION: Unit 2 is at rated :iditions when the following occurs: Indicatio - rn Wide Range level indicator (263-113) on the 902-4 panel are lost. Indicatio- _ n Wide Range digital level indicator (263-1 12) on the 902-5 panel are lost. These symptoms i 4icate a loss of... A. Instrument B_.; B. Essential Setr, - Bus. C. 125 VDC pa-.-' 2B-l. D. 125 VDC RBE Distribution panel. Explanation: ANSWER:

a.

Instrume-: Bus. Explanation: These level instrume-nts on the 5 panel are powered from the Instrument Bus.

Reference:

SDM *6000 Question Pedigree: BANK 21600S01 II Cog level: Memor. Rev I 03/28/02 1 of I Exam Submittal

Dresden 2002 ILT NRC Fxam QI/ Exam System K/A RO S RO 25 BOTH 218000 1(3.01 4.4 4.4 Automatic Depressurization Systemr Objective:2 I 81,-S 1-12 Knowledge of the effect that a loss or Restoration of reactor water level after a break that does malfunction of the AUTOMATIC not depressurize the reactor when required DEPRESSURIZATION SYSTEM will have on following: At 10:44 the following conditions exist on Unit 2: A stearn line break occurred in the drywell The MSIVs are closed. HPCI is operating and injecting into the vessel. Reactor water level is -45" and trending down at two inches per minute. Reactor pressure is 900 psig and steady. Drywell pressure is 1.5 psig and trending up at 0.2 psig per minute. At 10:47 the following 902-3 panel annunciators alarm: "* "ADS PERMISSIVE DW PRESS HI" (E-15) "* "ADS TIMER START" (B-13) "LPCI/CS PP AT PRESS" (H-13) "* "ADS INHIBIT" (G-I 1) What is the state the reactor at 10:54? A. Reactor water level is still trending down. B. HPCI is now maintaining level in the vessel. C. All five relief valves are open as required by ADS actuation. D. LPCI AND Core Spray pumps are running and injecting water into the vessel. Explanation: ANSWER

a.

Reactor water level is still trending down. Explanation: No conditions have changed that would allow HPICI to maintain level. The "ADS INHIBIT" is preventing the ADS valves from opening and with pressure steady the Low presspre ECCS pumps can not inject water to the vessel.

Reference:

DAN 902(3) BI13, E-15, G-11, and H-13, Question Pedigree: 218L-S I NEW Cog level: High Rev 2 I of I 03/28/02 Exam Submittal

Dresden 2002 [1,1 NRC Exam Q# 26 Exam 30T] I System 218000 K !A 1(4.01 RO 3.7 Automatic Depressurization Systemr Objective: 2 18L-S 1-05 Knowledge of AUTOMATIC Prevent inadvertent initiatior of ADS i DEPRESSURIZATION SYSTEM design feature(s) and/or interlocks which provide for the following: The Reactor is operating at 70% power when a LOCA condition develops in tile Drywell. The following is a timeline of ADS associated events: 17:15:00, Division I, High Drywell Pressure 17:15:30, Division I, Low Low Reactor Water Level 17:15:35, Division 11, ECCS Discharge Permissive 17:15:55, Division 1I, Low Low Reactor Water Level 17:16:01, Division 1, ECCS Discharge Permissive At what time will the FIRST 120 second Automatic Depressurization time delay time-out? A. 17:17:30 B. 17:17:35 C. 17:17:55 D. 17:18:01 Explanation: ANSWER:

a.

17:17:30 EXPLANATION: Because it only takes a simultaneous Divisional signal of High Drywell Pressure and Low Lcv Reactor Water Level to start the logic. There are two 120 second timers. Each requires that its respective Division I and Division II :ontacts for High Drywell Pressure and Low Low Reactor Water Level close in order to make-up the circuit to -he 120 second timer. >100 psig ECCS discharge pressure is not required in the circuit to initiate the 120 sec.:.nd timer. Keference: 5UM 2,1800 Cog level: High 03/28/02 Question Pedigree: Modified 21800S0211 -v 1 I of I Exam Submittal SRO 3.9

Jen 2002 II, 1"NRC Ixam QExam I

<K/A RO SRO 27 BOTHr --)00 A1.02 3.5 3 RHR/LPCI: Torus/Suppression Pool L :.:'ing Mode Objective: 203L-SI-3 Ability to predict and/or monitor chan.-i ;n System flow parameters associated with operating t:" RHR/LPCI: TORUS/SUPPRESSION 7*OL COOLING MODE controls including During torus cooling, after LPCI pump , has stabilized, the LPCI / CCSW Heat Exchanger should have... A. 3500 gpm of LPCI flow for each L---[ pump. B. 5000 gpm of CCSW flow for each --SW pump. C. a differential pressure of 20 psid x-LPCI system pressure greater than CCSW pressure. D. a differential pressure of 20 psid wi:- CCSW pressure greater than LPCI system pressure. Explanation: ANSWER:

d.

at a differential pressure of 20 p-:A with CCSW pressure greater than LPCI system pressure. When the LPCI pump is started CCSW p-7ssure is much greater than LPCI pressure. After LPCI flow stabilizes CCSW pressure should be 20 p-:z greater than LPCI pressure.

Reference:

DOP 1500-2 Question Pedigree: BANK 21900S0021 Cog level: Memory Rev 1 03/28/02 1 of I Exam Submittal

Dresden 2002 ILT NRC Exam System 219000 K/A A2.05 RHR/LPCI: Torus/Suppression Pool Cooling Mode Objective: 20.3L-S 1-6 Ability to (a) predict the impacts of the following on the RIIR/LPCI: TORUS/SUPPRESSION POOL COOLING MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A.C. electrical failures Unit 2 was at power with the Unit 2 Isolation Condenser OOS, when a transient occurred. The following conditions exist: Bus 24 was damaged by a fire and has been taken out of service. All rods are inserted. MSIVs are closed to conserve inventory. RPV level has remained at +10 inches. HPCI is being used for reactor pressure control. The (1) CCSW pumps can be started and Torus cooling can be performed by (2) 1 2 A. 2A and 2B each CCSW pump supplying a different LPCI/CCSW HX B. 2A and 2B both CCSW pumps supplying one LPCI/CCSW HX and utilizing the cross-tie line C. 2C and 2D each CCSW pump supplying a different LPCI/CCSW HX D. 2C and 2D both CCSW pumps supplying one LPCI/CCSW HX and utilizing the cross-tie line Explanation: ANSWER:

b. 2A and 2B and both CCSW pumps supplying one LPCI/CCSW HX and utilizing the cross tie line.

EXPLANATION: The CCSW pumps that have power available, are on one bus, in one division, and therefore cannot be lined up one to each HX.

Reference:

SDM 277000, SDM 203000 and DOP 1500-Question Pedigree: 02 Modified 27700S0321 Cog level: High Rev I 03/28/02 I of I Exam SumomittaI Exam RO RO 3.3 SRO 3.5 I ObJective: 203L-St-6

Dresden 2002 fILT NRC Exam F/xain BOTHF Systeni "223001 Primary Containment System and Auxiliaries Ability to monitor automatic operations of [hle PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES including: Unit 3 is at rated conditions with the follnwinc. Drywell pressure is 1.0 psig. Torus pressure is 1.5 psig. Reactor Building pressure is -0.25 inches of water (0 psig). 03/28/02 I of I Exam Submittal QI/ 29 K/A A3.02 RO 3.4 SRO 3.4

Dresden 2002 ILT NRC ExI: Q4 Exam System K A RO SRO 30 SRO 223002 2.4.4 4.0 4.3 Primary Containment Isolation System/Nuclear Steam 1: Ob*.l- .e: Supply Shut-Off Emergency Procedures and Plan Ability to reco.:- _-e abnormal indications for system operatingc para-7 :---;'s which are entry-level conditions foi emerg~ency aný --"ormial oper-atinigprocedur-es. Given the following conditions: "* HPCI test is in progress in accordance with DOS 23(+'.- :nd is operating at full flow through the test line with suction from the CST. "* A small steam leak develops in the HPCI room. "* Temperature in the room is 210'F and increasing at 5"7 -,r minute. "* The HPCI system flow remains constant. As a result of this the... A. reactor must be shutdown. B. steam supply to the HPCI system must be isolated. C. HPCI system will isolate when room temperature reaches 3007F D. HPCI room cooler will trip and Standby Gas Treatment will au-: -zart. Explanation: ANSWER: b, steam supply to the HPCI system must be isolated. Explanation: SRO only criteria 5 The isolation of HPCI should have occurred at 200°F. DEOP 300-1 m,-tds to be entered and the correct actions that need to be taken are "isolate all discharges into affected areas". L cram is not needed until another temperature reaches max safe. Standby gas will be running prior to s:r'- the surveillance and these conditions will not trip the room cooler.

Reference:

SDM 223005, SDM 206000, and DEOP Question.- digree: 300-1 NEW Cog level: High. Rev 1 03/28/02 1 of I Exam Submittal

Dresden 2002 II,T NRC l'axam Q#/ E x, System 1K/A R O SRO 31 BOTH 223002 K 1. 19 2.7 2.9 Primary Containment iso'lation System/Nuclear Steamr Objective: 208L-S 1-6 Supply Shut-Off Knowledge of the phs_':cal connections and/or Component cooling water systems cause-effect relationsýh :s between PCIS/NSSSS and the following: Given the following inormnation: A Loss of Coolant Accident (LOCA) has occurred inside the drywell. a The break has also caused a rupture in the RBCCW supply line resulting in a loss of RBCCW flow and pressure. The drywell atmosphere is prevented from entering the Reactor Building through the RBCCW system by.... A. the RBCCW expansýion tank. B. check valves in the RBCCW piping. C. manually isolating :he RBCCW systemn at the 923-1 panel. D. an automatic isolation by the Primary Containment Isolation Systemn. Explanation: ANSWER

c. manually isolating the RBCCW system at the 923-1 panel.

Reference:

Question Pedigree: SDM208000 Modified 20800S0081 Cog level: Memory Rev I 03/28/02 1 of I Exam Submittal

Dresden 2002 ILT NRC Exam Exam BOThI System 226001 Exam Submittal Q2 32 K/A K2.02 RO 2.9 SRO 2.9 03/28/02 I of I

Dresden 2002 ILT NRC lFxun1 System 226001 K/A K15.02 RHR/LPCI: Containment Spray System Mode Knowledge of the operational implications of the following concepts as they apply to RHR/LPCI: CONTAINMENT SPRAY SYSTEM MODE: Objective: 209L-S1-12 Water hammer The reason the LPCI pump discharge piping has a keep fill system is to... A. provide an indication of the LPCI pump integrity. B. minimize corrosion in the Torus and Drywell spray lines. C. prevent the LPCI/CCSW heat exchanger from becoming air bound. D. minimize the effects of water hammer on the system and pipe hangers. Explanation:

c.

minimize the effects of water hammer on the system and pipe hangers. The ITS bases states the reason for a ECCS keepfill system is to ensure rapid delivery of water to t-.- RPV and to minimize the effect of water hammer.

Reference:

SDM 203000 and ITS Bases 3.5.1 Cog level: Memory 03/28/02 Exam Submittal Question Pedigree: Modified 29900S0241 Rev 2 I of I Exam BOTH RO 2.6

l)resde- _' )02 1I LT NRC 'xam Q# FPxam Systew K/A RO SRO 34 BOTI1 23000( 1"6.05 3.3 3.4 RHR/LPCI: Torus/Suppression Pool Spray.' :e Objective: r03')3L-S1-12 Knowledge of the effect that a loss or Suppression pool malfunction of the following will have on the RHR/LPCI: TORUS/SUPPRESSION POOL SPRAY MODE: One hour ago a small LOCA developed on U.-. 2 and tile following conditions exists: RPV level is 30 inches and stable. Drywell pressure is 2.5 psig and stead% Drywell temperature is 210'F and trerf '-ig down slowly. LPCI is being used for Torus cooling..-d Aspray. Torus level is 14 feet and rising slowl k Subsequently: Discharge pressure and flow on the LPS-[ pumps started fluctuating The following annunciators alarm: "2A LPCI HDR FLOW LOW" "2B LPCI HDR FLOW LOW" This indicates that...... A. The ECCS keepfill pump has tripped. B. The LPCI pump suctions have auto swapped over to the CST. C. The ECCS ring header suction strainers ha\\ e become clogged. D. The EDG have started and are now supplying power to the LPCI pumps. Explanation: ANSWER

c. The ECCS ring header suction strainers ha\\ become clogged.

Explanation: With the LPCI pump running a loss of keepfill 7-imp will not give these indications. The LPCI suctions do not swap over to the CST without operator action a.- the LPCI pumps are not being powered from the EDGs.

Reference:

SDM203000, 209001 and NRC bul]erin 93-Question Pedigree: 02 NEW Cog level: High Rev I 03/28/02 I of I Exam Subm ittal

Dresden 2002 IIT NRC Exam System 234000 K/A 2.1.2 Fuel 1Handling Equipment Conduct of Operations Objective: 234L003 Knowledge of operator responsibilities during all modes of plant operation. Fuel movements are being performed in the Unit 2 core. The Unit 2 NSO and Fuel Handlers cannot agree as to which move is to be performed next. According to Unit 2 Master Refueling Procedure, DFP 800-1, which ONE of the following personnel is to be contacted? A. Unit 3 NSO A. Unit 3 NSO B. Fuel Handling Foreman C. Nuclear Materials Custodian D. Control Room Nuclear Observer Explanation: ANSWER J. Control Room Nuclear Observer Reference DFP 0800-1 Question Pedigree: Bank 23400S01 I Cog level: Memory Rev 1 "28/02 I of I Exam Subm -*a Q4 Exam 130TI RO 3.0 SRO 4.0 A. Unit3 NSO

Dresden 2002 111, NRC F, am C. Q# 36 FXam BOTh I System 239001 K/A K5.06 RO 2.8 SRO 2.9 Main and Reheat Steam System ObJective: 239L-S ]- 12 Knowledge of the operational implications of the Air operated MSI V's following concepts as they apply to MAIN AND REHEAT STEAM SYSTEM: Unit 3 is at rated conditions when the following occur: Instrument Air header pressure rapidly goes to 0 psig. This would result in.. A. the FRVs failing closed. B. the inboard MSIVs starting to close. C. the outboard MSIVs starting to close. D. the FRVs immediately locking up in their current position. Explanation: ANSWER:

c.

The outboard MSIVs starting to close. Explanation: The outboard MSIVs are held opened by Instrument Air. The inboard MSIVs are held open by drywell pneumatics. The FRV have a backup air supply that lasts for 75 minutes then they "lockup"

Reference:

Question Pedigree: DOA 4700-1 Modified 03000S0541 Cog level: High Rev I 03/28/02 I of I Exam Submittal Main and Reheat Steam System Objective:. ")"_o9L-S 1-12 Knowledge of the operational implications of the Air operated MSIV's following concepts as they apply to MAIN AND REHEAT STEAM SYSTEM: Unit 3 is at rated conditions when the following occur: Instrument Air header pressure rapidly goes to 0 psig. This would result in..

Dresden 2002 I. F NRC Exam Q Exam System K/A RO %RO 37 RO 241000 2.4.4 4.0 4.3 Reactor/Turbine Pressure Regulating System Objecti\\z -:41L-SI-6 Emergency Procedures and Plan Ability to recognize --iormal indications for system operating parameterz liich are entry-level conditions for emergency and abn. --._fl operating procedures. Unit 2 is at rated condition when a failure of the EHC system causes atr:- -ient to occur. As a result of the transient: RPV level dropped to 27 inches and is now rising. RPV pressure stabilized at 1070 psig. With these indications the operating team should enter.... A. DEOP 100-1, RPV Control, due to RPV level. B. DEOP 100-1, RPV Control, due to RPV pressure. C. DOA 600-1, Transient Level Control, due to RPV level. D. DOA 5650-03, Turbine Control Valve or Bypass Valve Failed Open. :-e to reactor pressure. Explanation: ANSWER

b. DEOP 100-1, RPV Control, due to RPV pressure.

Explanation: Entry for DEOP 100-I is 1060 psig and 8 inches. Entry for DOA 600-1 is inches.

Reference:

DEOP 100-1, DOA 600-1, and DOA 5650-Question Pedigree: 03 NEW Cog level: High Rev I 03/28/02 1 of I Exam Submittal

Dresden 2002 ILF NRC Exam QI# Exa;- System K/A R(O SRO 38 BOTZ 245000 K3.02 3.9 1.0 Main Turbine Generator - Auxiliary Systems Objective: 245L-S1-6 Knowledge of the effec: -:t a loss or Reactor pressure malfunction of the MAP. TURBINE GENERATOR AND AL_ -LIARY SYSTEMS will have on following: Given the following co.n- -:ons on Unit 3 Reactor Power i o -5` The Main Turbir. '.lain Generator is on line. A fault occurs that causes :.e Main Generator field breaker to open. Which of the following o-u:,r? A. A load reject scram I-:urs. B. Reactor pressure go- -own. C. Reactor scrams on hn - pressure. D. The bypass valves co-.iol reactor pressure. Explanation: ANSWER:

d.

The bypass vale-x control reactor pressure. EXPLANATION: With reactor power less th:z 45% power the load reject scram is bypassed. Reactor pressure is controlled by the bypass valves and prestzure will not go down or cause a reactor scram.

Reference:

SDM212001 a:. 1 DOA 5600-I Question Pedigree: New Cog level: High Rev I 03/28/02 I of I Exam Submittal

Dresden 2002 II,'fNRC Exam Q.: 39 Exam RO System 256000 K/A A2.12 RO 3.1 SRO 3.1 Reactor Condensate System Objective: 274L-S 1-l2 Ability to (a) predict the impacts of the Loss of equipment component cooling water systems following on the REACTOR CONDENSATE SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Both Units are operating at rated conditions with the following pumps running: 2A TBCCW 3A TBCCW The following then occurs: Annunciator "U2 or U3 TBCCW PRESS LOW" alarms. U2 TBCCW DISCH HDR PRESS indicates 0 psig. What is the impact and how is the situation controlled? IMPACT HOW CONTROLLED A. Cooling is lost the Condensate Pump seal coolers. Start the 2B TBCCW pump only. B. Cooling is lost the Condensate Pump seal coolers. Start the 2B OR the 3B TBCCW pump. C. Cooling is lost to the MG Set oil coolers. Start the 2B TBCCW pump only. D. Cooling is lost to the MG Set oil coolers. Start the 2B OR the 3B TBCCW pump. Explanation: ANSWER

a. Cooling is lost the Condensate seal coolers, and Start the 2B TBCCW pump.

Give the students a copy of the DANs for loss of pump and low expansion tank with Section B.3.e removed or blanked out. (This is the section that lists the loads in the loss of pump DAN)

Reference:

Question Pedigree: DAN 923-1 C-2 and D-2 NEW Cog level: High Rev I 03/28/02 1 of[ Exam Submittal

Dresden 2002 I I.F NRC Exam Exam RO System 259001 K/A 2.1.2 RO 3.0 SRO 4.0 Reactor Feedwater Systemr Objective: 2590 1LK006 Conduct of Operations Knowledge of operator responsibilities during all - des of Condct f Opratonsplant operation. Unit z is operauing at rateu power wnen multiple reed Kegulating Station high vibration alarms are re.:-.-ed and feedwater flow oscillations are observed. The operator is to maintain feedwater flow for above.

2) inches.

(1) seconds, OR until reactor level is restore: 1 2 1 2 A. 15 15 B. 15 20 C. 60 15 D. 60 20 Explanation: ANSWER:

c.

60, 15

Reference:

DOA 3200-01 Question Pedigree: Bank 25901 S0091 Cog level: Memory Rev 2 03/28/02 I of I Exam Submittal Q/1 40

Dresden 20(ý -. T NRC Exam Q# Exam System K/A RO SRO 41 BOTH 259002 A2.01 3.3 3.4 Reactor Water Level Control System Objective: 259L-S-6 Ability to (a) predict the impacts of the -:s of any number of main steam flow inputs following on the REACTOR WATER LEVEL CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: While operating at rated conditions, which of the t\\oi :in-signals will cause the Un-it 2 FWLC ýsystem to transfer fr-om 3 -Element to I - Element control, ANE iat action should tile operator take? SIGNAL ACTION A. 2A Feed Flow instrument fails "BAD Take manual control of the FRVs QUALITY" B. 2A Steam Flow instrument fails "BAD Depress the "I-ELEM" pushbutton QUALITY" C. "A" NR level instrument fails to "BAD Take manual control of the FRVs QUALITY". D. "A" NR level instrument fails to "BAD Depress the "I-ELEM" pushbutton QUALITY". Explanation: ANSWER:

b.

2A Steam Flow instrument fails "BAD QUALIT'.-. Depress the "I-ELEM" pushbutton. Micslnfo: The 2A Feed Flow instrument will cause the switch to :ccur, but going to manual control of the FRV is not the correct action. The level input will switch to another ir-ut device. The operator needs to depress the "I ELEM" push button to have indication match actual ccrditions. Taking manual control of the FRV actually would complicate the situation.

Reference:

SDM 259002 and DAN 902-5 G-8 I Question Pedigree: Modified 25902S0391 Cog level: High Rev I 03/28/02 1 of Exam Submittal

l)resden 2002 IL[" NRC Fxam Q# Exam System K/A RO SRO 42 BOTH 259002 A4.06 3.1 3.2 Reactor Water Level Control S\\ stem Objective: 2591,-S2-8 Ability to manually operate and or monitor in DP/Single/three element control selector switch: Plant the control room: Specific The following indications are present for the FWLC system. "I-ELEM" is white "* "AUTO" is amber "3-ELEM" is flashing amber These are an indication that the operator selected.... A. 3 Element control and the s\\ stem is still operating in 3 Element control. B. 3 Element control and the system automatically switched to I Element contrrol. C. I Element control and the system is still operating in I Element control. D. 1 Element control and the system automatically switched to 3 Element contr 03/28/02 Exam Submittal ol. I of I Explanation: ANSWER:

b.

3 Element control and the system automatically switched to I Element control. Explanation: White indicates FWLC is currently in this mode. Amber indicates was in this mode and has switched. Flashing amber indicates no longer available for use.

Reference:

SDM259002 Question Pedigree: NEW Cog level: High Rev I rol.

l)resden 2002 II T NRC Exam QExam System KiA RO SRO 43 SRO 215005 2.2.22 3.4 4.1 Average Power Range Monitor/Local Power Range Monitor Objective: 2 15L-S5-7 System5 Equipment Control Knowledge of limiting conditions for operations and safety limits. Unit 3 is in coastdown in preparation for a refueling outage with the following conditions: Reactor power is 50% Total core flow is 85% Flow Converter #1 is downscale due to Instrument Maintenance performing calibration. No LCO actions are in effect. Flow Converter #2 fails to 102%. Which, if any, of the APRMs are now considered INOPERABLE? A. APRMs 1, 2, and 3 B. APRMs 4,5, and 6 C. Ail APRMs D. None of the APRMs Explanation: ANSWER

b. APRMs 4, 5, and 6 ITS Bases 3.3.1.1.2.b states that an APRM flow converter is considered inoperable whenever it cannot deliver a flow signal less than or equal to actual recirculation flow conditions for all steady state and transient conditions while in MODE 1. Reduced flow or downscale flow converter conditions due to planned maintenance or testing activities during derated plant conditions will result in conservative setpoints for the APRM flow bias functions, thus maintaining the function operable.

Flow converter #2 provides input to APRMs 4,5, and 6. These are now considered inoperable due to them having a higher flow signal than actually exist. Flow converter #1 provides input to APRMs 1, 2, and3. These are still OPERABLE since they are downscale resulting in a more conservative setpoint.

Reference:

Question Pedigreeo ITS Bases 3.3.1.1.2.b and SDM 215005 New Cog level: High Rev 2 03/28/02 1 of I Exam Submittal

Dresden 2002 ILT NRC Exam Q// Exam System K/A RO SRO 44 BOTH 261000 A2.04 22.7 Standby Gas Treatment System Objective: 261 L-S I-. Ability to (a) predict the impacts of the following on High train moisture content the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: The following condition exist on Unit 2: 2/3A SBGT is running due to an auto initiation. 2/3B SBGT is in STBY Then the following occurs: "STBY GAS TRT SYS A TROUBLE" annunciator alarms. The 2/3A AIR HEATERS indicate OFF. What is the potential problem and what action should the operator take to correct the _-: blem? Problem A. Moisture could enter the charcoal, which decreases the charcoal filtration efficiency. B. The charcoal is NOT warm enough to adsorb the radioactive iodine. Acti,: Verify the 2/3B SBGT starts Verify the 2/3B SBGT star-ts C. Moisture could enter the charcoal, which Reenergize the heaters on the 2. 3A SBGT decreases the charcoal filtration efficiency. D. The charcoal is NOT warm enough to adsorb the Reenergize the heaters on tht 1 3A SBGT radioactive iodine. Explanation: ANSWER:

a.

Moisture could enter the charcoal, which decreases the charcoal efficiency and ve--*,' the 2/3B SBGT starts. Explanation: If the heaters are not energized moisture is not removed and the efficiency of the charc:1-1 decreases. The actions that are required by the DAN under these conditions are verify the 2/3 B SBGT :-arts.

Reference:

SDM 261000 and DAN 923-5 A-6 Question Pedigree: New Cog level: High 03/28/02 I of I Exam Submittal Rev I 7 ----------

Dresden 2002,1l F NRC "Axam System 261000 K2.\\ \\2. 13 Standby Gas Treatment System Ability to (a) predict the impacts c -he following on the STANDBY GAS TREATMENT SYSTEM; and (b1 - sed on those predictions, use procedures:- zorrect, control, or mitigate the consequenlle-c of those abnormal conditions or operationis 03/28/02 I Objective: 261 L-S 1-07; 261 00LKOO2 I of I Exam Submittal Q#i 45 Exam RO RO 3.4 SRO 3.7 IIigh secondary containment ventilation exhaust radiation The following plant conditions ex'_: Unit 2 is at 820 MWe. Unit 3 is off-loading fuel.: r 550 days of operation. 2/3A SBGT selector switck :s in PRI 2/313 SBGT selector switch -s in STBY The following then occurs: Reactor Building Ventila:. :n Exhaust Duct trends up to 7 tmrero/hr. Which of the following describes t-2 system response and the actions required if this does NOT occur? SYSTEM RESPONSE ACTION REQUIRED A. 2/3A SBGT starts and 2/3B SBET starts if the Immediately suspend all irradiated fuel moves. 2/3A SBGT fails to start. B. 2/3A SBGT starts and 2/3B SBGT starts if the Restart Reactor Building ventilation. 2/3A SBGT fails to start. C. 2/3B SBGT starts and 2/3A SBGT starts if the Immediately suspend all irradiated fuel moves. 2/313 SBGT fails to start. D. 2/3B SBGT starts and 2/3A SBGT starts if the Restart Reactor Building ventilation. 2/313 SBGT fails to start. Explanation: ANSWER: a 2/3A SBGT starts and 2/313 SBGT starts if the 2/3A SBGT fails to start. Immediately suspend all irradiated fuel moves. EXPLANATION: c and d are wrong because the SBG7 train in PRI starts and the SBGT train in STBY starts if the PRI fails to start b is wrong because on a valid :?,actor building ventilation radiation isolation signal reactor building ventilation is not restarted.

Reference:

DOA 7500-01 and DAN 923-5 A-6 Question Pedigree: Modified 26100S0191 Cog level: High. Rev I

Dresden 2002 ILT NRC Exam Qg Exam System K/A RO SRO 46 RO 262001 A1.01 3.1 3.4 A.C. Electrical Distribution Oblective:262L-S 1-06 Ability to predict and/or monitor changes in Effect on instrumentation and controls of switching power parameters associated with operating the A.C. supplies ELECTRICAL DISTRIBUTION controls including: Given the following conditions: "BUS 24 OVERCURRENT" annunciator alarms and the plant responds as expected. The Unit 2 Diesel Generator fails to start. What is the affect, if any, on the LPCI injection valves (1501-2 1A/B and 22A/B) during this event? A. remain unaffected. B. lose valve power for 20 seconds. C. the valves will open until power is restored. D. lose valve power indefinitely until restored manually. Explanation: ANSWER:

b.

lose valve power for 20 seconds. Explanation: In the condition described above the valve lose power when Bus 24 is lost and the DG fails to start. MCC 28 7/29-7 has and auto swapping feature that activates once MCC 29-7 loses power. 17 seconds after power is lost MCC 28-7/29-7 will be powered from Bus 28.

Reference:

203L-S1, SDM262001 Question Pedigree: BANK 26201S0116 Cog level: High Rev I 03/28/02 1 of I Exam Stu: iItta

Dresden 2002 [LT NRC Exam E-xarn BOT[0I System 262001 Exam Submittal Q4 47 K/A K2.01 RO 3.3 SRO 3.6 03/28/02 I of I

Dresden 2002 LTI NRC F\\ Exam BOTHI Exam Submittal Q# 48 System 262002 1K i,. 1,3. V 1,O 2.9 S RO 3.1 Uninterruptable Power Supply (A.C./D.C.) i () tive: 262L-S2-12 Knowledge of the effect that a loss or Process m owi:-g Plant-Specific malfunction of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) will have on following: Unit 2 just completed a refueling outage. Start up is in progress. The Mechanical Vacuum pump is in operation. "A" Main Steam Line radiation monitor has failed upscale Then the following occurs: "* "ESS UPS TROUBLE" annunciator alarm. "* "ESS UPS ON DC OR ALTERNATE AC" annunciator aL--n. "120/240V ESS BUS VOLT LO" annunciator alarm. The NLO reports from the AEER that the ESS Bus voltage zero. As a result of this the.... A. mechanical vacuum pump trips. B. refuel floor rad monitors fail downscale. C. 24/48 VDC system battery chargers lose power. D. Reactor Feed Pump minimum flow valve fails closed. Explanation: ANSWER

a. The mechanical vacuum pump trips Explanation The main steam line rad monitors are powered from the ESS bus ar.- when the ESS bus loses power the rad monitors lose power, which trips the mechanical vacuum pump.

Reference:

SDM262006 and 272002 Questio-Pedigree: NEW Cog level: High Rev 2 03/28/02 I of]

Dresden 2002 ILT NRC Exam System 264000 A 1.09 Emergency Generate:-, Diesel/Jet) Ability to predict an, -- monitor changes in parameters associate- '. ith operating the EMERGENCY GEN RE ZATORS (DIESEL/JET) controls including: Objective: 264I-S 1-6 Maintaining minimum load on emergency generator (to prevent reverse power) The NSO is performi:., a EDG surveillance: To prevent a reverse r ý:,*er trip of the output breaker, after closing the output breaker the NSO must (I)_ load -.:ing the _ (2) control switch in accordance with DOS 6600-01, Diesel Generator SurveillanI-Tests. 2 -I _________________________________________ GOVERNOR I A. raise B. raise VOLTAGE REG C. lower GOVERNOR VOLTAGE REG I of I D. lower 03/28/02 Exam Submittal Q# 49 F' !:i BI-I7 RO 3.0 SRO

3.1 Explanation

a.

raise, GOVERNOR Explanation: By raising the Governcr control switch the DG accepts some load to prevent a reverse power condition.

Reference:

DOS 6600-) 1, Diesel Generator Question Pedigree: Surveillance Tests. New Cog level: Memory Rev 2

Dresden 2002 IIT NRC E'xam Exam System K/A RO SRO 50 RO 264000 A3.06 3.1 3.2 Emergency Generators (Diesel/Jet) ObJective: 264L-S 1-06 Ability to monitor automatic operations of the Cooling water system operation EMERGENCY GENERATORS (DIESEL/JET) including: Given the following: A Loss of Coolant Accident has occurred on Unit 2 resulting in drywell pressure of I I psig. Unit 2 Emergency Diesel Generator is supplying power to the emergency bus. A large leak develops in the Unit 2 DG Cooling Water System. With NO operator action, which statement below describes the response of the Unit 2 EDG to this condition? A. Unit 2 EDG continues to run to destruction. B. Unit 2 EDG trips when cooling water temperature reaches 200'F. C. Unit 2 EDG trips when cooling water pressure drops below 35 psig. D. When cooling water pressure drops to 35 psig, the Unit 2 EDG will continue to run for 6 minutes and then shutdown. Explanation: ANSWER:

a. Unit 2 EDG continues to run to destruction.

Explanation: The student must recognize the a high drywell signal causes the engine trips to be bypassed. With trips bypassed and loss of the heat sink, the engine will run to destruction.

Reference:

SDM 264001 Question Pedigree: BANK 26400SO261 Cog level: High Rev 2 03/28/02 1of I Exam Submittal

Dresden 2002 I[,T NRC Fxam Exam BOTIH Systc m 271000 K5.A 1K-5.07 RO

2.7 Explanation

ANSWER

d. to allow for the decay of gaseous radioactive nuclides to particulate Explanation The recombiner recombines hydrogen and oxygen. The is no mixing in the holdup pipe. And absorp::mn occurs in the charcoal beds which allows for the decay to take place. The Flux Tlit monitor is not bypassed.,,. hen the adsorbers are.

Reference:

UIV2 / 7 U10 IQuestion Pedigree: Modifie K/A for Cog level: Memory d Question #23 of the Aug 97 NRC Exam that exam was 271000K508 Rev I I of I Exam Submitta) Q/1 51 SR, I2.) 03/28/02 -1

Drcsde:, 12 Il TNRC [`xam Svstenc 28800( K/A A2.0l I iýant ventiation zysterns Ability to (a) predict the impacts of the following on the PLANT VENTILATION SYSTEMS; and (b) based on those predictior: use procedures to correct, control, or mitigate consequences of those abnormal conditions o operations: Objective: 288L-SI-12 High drywell pressure: Plant-Specific Unit 3 is at rated conditions when the followii-.- _ccur: A small LOCA develops inside the dr.- Ill Drywell pressure is 1.5 psig and risin* !wly What will be the impact as drywell pressure c:.-finues to rise and what actions are necessary? IMPACT ACTION A. Reactor Building ventilation isolates Restart Reactor Building ventilation B. Reactor Building ventilation isolates Verify SBGT system operating C. Turbine Building ventilation isolates Restart Turbine Building ventilation D. Turbine Building ventilation isolates Verify SBGT system operating 03/28/02 I of I Exam Submittal Q5 52 Exam BOTH R() SRO 3.4 Explanation: ANSWER:

b.

Reactor Building ventilation isolates and \\ trify SBGT system operating Explanation: The turbine building ventilation will not isolate Ind restarting the Reactor Building ventilation is not the correct action..

Reference:

Question Pedigree: DAN 902-5 G5 and 923-5 A l New Cog level: High Rev I

l)resden 2002 IlT'F NRC Exam Examn SRO System 290002 K1A 2.1.32 RO 3.4 SRO 3.8 Reactor Vessel Internals Objective: 2021-S 1-7 Conduct of Operations Ability to explain and apply system limits and precautions. Prior to returning to two loop operation fiom one loop operation which of the following limits must be met and what is the reason for that limit? LIMIT A. The temperature difference between the bottom head coolant and the recirc loop coolant in the loop to be started is < 145°F. REASON To prevent a violation of the RPV pressure and temperature limitation that minimize the chances of brittle fracture from occurring. B. The temperature difference between the recirc To prevent a violation of the RPV pressure and loop coolant in the loop to be started and the temperature limitation that minimize the chances of reactor vessel coolant is < 50'F. brittle fracture from occurring. C. The temperature difference between the bottom To prevent damage to the fuel cladding that would head coolant and the recirc loop coolant in the result from the sudden increase in power due to the loop to be started is < 145 0F. injection of cold water. D. The temperature difference between the recirc To prevent damage to the fuel cladding that would loop coolant in the loop to be started and the result from the sudden increase in power due to the reactor vessel coolant is < 50'F. injection of cold water. ExplanaLlon: ANSWER:

b. The temperature difference between the recirc loop coolant in the loop to be started and the reactor vessel coolant is < 50'F. To prevent a violation of the RPV pressure and temperature limitation that minimize the chances of brittle fracture from occurring.

Explanation: A and C are wrong because the other limit is between the operating loop and the bottom head coolant temperature. D is wrong because the limit is for brittle fracture reasons, the thermal limits, MCPR, APLHGR, and LHGR prevent damage to the cladding. ITS sections 3.4.9 and 3.2 explain the reason for these limits. ')3/28/02 I of I Exam Subm -taI QfI i

Reference:

DOP 0202-01, ITS section 3.4,9 and 3.2 Question Pedigree: New Cog level: High Rev I

I)resden 2002 IfLT NRC r'xam Qt' Exam System K/A RO, SRO 54 BOTH1 288000 2.1.33 3.4 4.0 Plant Ventilation Systems Objective: 288L-S1-5 Conduct of Operations Ability to recognize indications for system operating parameters which are entry-level conditions for technical specifications. Unit 3 is at rated conditions. Which ONE of the following would require LCO action? A. Reactor Steam Dome Pressure is 900 psi~g. B. Secondary Containment at -0.1 inches H20. C. Reactor Coolant System identified leakage is 2 gpm. D. One of the Turbine Building to Reactor Building Interlock Doors is closed but unable to be opened. Explanation: Answer:

b. Secondary Containment at -0.1 inches H20.

Explanation: The other items are not LCO entries. Steam dome pressure must be between 785 and 1005 psig, Unidentified leakage is allowed to go up to 5 gpm. The Turbine Building to Reactor Building Interlock Doors must be closed but do not have to be able to be opened.

Reference:

ITS 3.6.4.1 Question Pedigree: New Cog level: Memory Rev I 03/28/02 I of I Exam Submittal

Dresden 2002 ILT NRC Exam Q# Exam System K/A RO0 SRO 55 RO 290003 A3.01 3.5 Control Room HVAC Object 88L-$3-05 Ability to monitor automatic operations of the Initiation/reconfie: :n CONTROL ROOM I IVAC including: Given the following information regarding the Control Room Ventilat! -.xstcm: Train "A" Air Handling Unit control switch has a RED-TARG-: Train "B" Air Handling Unit is operating. Air Filtration Unit is secured. The Train "B" Air Handling Unit Isolation Dampers (XCV-2 3----' 1-059 A and B) are OPEN All other dampers are CLOSED. The Control Room Ventilation system is in the mode. A. smoke purge B. normal operating C. emergency operating D. isolation/recirculation Explanation: ANSWER:

d.

isolation/recirculation Explanation: With all dampers closed except the 59 A and B, and the B AHU the recir:,: -he air in the Control Room with the AFU secured.

Reference:

SDM 288003 Question Ped:----,e: BANK 2880>.. :41 Cog level: High Rev I 03/28/02 1 of I Exam Submittal

Dresden 2002 I[." UNRC Fxam Emir SRO System 295001 Exam Submittal 56 K/A 2.2.25 RO 2.5 SRO 3.7 Partial or Complete Losz- -"Forced Core Flow Circulation IObjective: 202L-S 1-7 Equipment Control Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. What is the bases for thý '-CO that states "Two recirculation loops with forced flow shall be in operation." A. Prevent entering the !istability Region" of core flow. B. To prevent excessixe,ibrations of the jet pump risers. C. Natural circulation \\ :! not remove the heat generated by the fuel. D. To ensure that the acqimptions of the LOCA analysis are satisfied. Explanation: ANSWER:

d. To ensure that the assLnptions of the LOCA analysis are satisfied.

Explanation: Two recirculation loops -e normally required to be in operation with their flows matched within the limits specified in S.R. 3.4.1.1 : ensure that during a LOCA caused by a break of piping of one recirculation loop the assumptions so the LOCA analysis are satisfied.

Reference:

ITS Bases 3.- 1 Question Pedigree: New Cog level: Memory Rev I 03/28/02 I of I

Dresden 2002 1[,T N RC Fxam Exam SRO System 295002 Exam Submittal K/A 2.4.49 RO 4.0 SRO 4.0 Loss of Main Condenser Vacuum IObjective: 275L-S 1-6 Emergency Procedures and Plan Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. Unit 2 is at rated conditions when the following occurs: Annunciator 902-7 B-15 "SCREEN WASH CONTROL PANEL TROUBLE" alarms The NLO reports the following: A large buildup of fish on the inlet side of the travelling screens. "* There is a 14 inch level difference across the travelling screens. 15 minutes later the following occurs: The NSO reports vacuum starting to trend down at 0.5 inches Hg per minute. The NLO reports the level difference is getting worse as more fish are accumulating on the travelling screens. Based on these reports, which of the following actions must be performed, AND what is the reason for the action? Action Reason A. Scram protect the condenser from over pressure and maintain heat sink available B. Lower power and leave only one Circulating maintain vacuum and CCSW system available Water pump running C. Scram maintain vacuum and CCSW system available D. Lower power and leave only one Circulating protect the condenser from over pressure and maintain Water pump running heat sink available QIP 57 03/28/02 Explanation: ANSWER: a Scram the protect the condenser from over pressure and maintain heat sink available. Explanation: SRO only per criteria I and 5. The DOA has under immediate actions that if a loss of condenser vacuum is IMMINENT and bar rack level difference does not improve scram the reactor. The Tech Spec Bases states that the reason for a low vacuum scram is to protect the main condenser from over pressure and maintain the heat sink available

Reference:

DOA 4400-06 and Tech Spec Bases 3.3.1.1 Question Pedigree: New Cog level: High Rev I I of I

Dresden 2002 I T N R C Exam Q4 58 Exam BOTHl System 295002 K/A AK2.04 RO 3.2 SRO 3.3 Loss of Main Condenser Vaculun le bjective: 245L-Sov-05 Knowledge of the interrelations between LOSS i Reactor/turbine pressure rgulating system OF MAIN CONDENSER VACUUM and the following: ANSWER A reactor cooldown is in progress on Unit 2 using the BYPASS VALVE-OPENING JACK. The circulating water pumTps trip. What will occur? A. The MSIV's will isolate on low pressure. B. The rupture disk on the LP turbine will blow out. C. The bypass valves will close on low main condenser vacuum. D. Turbine exhaust hood spray will initiate on high backpressure. Explanation: ANSWER:

c.

The bypass valves will close on low main condenser vacuum. Explanation: When the circ water pumps trip vacumm goes away. When vacuum is less than 7 inches of Hg the bypass valves go closed to protect the condenser.

Reference:

SDM 275001 and SDM 241000 Question Pedigree: Modified 24501S0401 Cog level: High Rev 2 03/28/02 Exam Submittal I of I

Dresden 2002 iN - Ni( Fxan Q11 Exam System K/A RO SRO 59 BOT I[ 295003 AA 1.03 4.4 4.4 Partial or Complete Loss of A.C. Power Objective: 264S LI-12 Ability to operate and/or monitor the following S-eoms necessary to assure safe plant shutdown as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Unit 2 is at 40% power when the following alarms a17. -nciate: 4KV MAIN FEED BKR TRIP 4KV BUS 23-1/24-1 VOLT LO 4KV BUS 24-I VOLTAGE DEGRADED Upon investigation, you notice: the Main Feed Breaker for Bus 24-1 is tripped Bus 24-1 is de-energized. Unit 2 Emergency Diesel Generator is NOT rr:ning. What actions, if any, are required? A. Be in cold shutdown condition within 7 days. B. Attempt to manually start the U2 Diesel Generater-from the 902-8 panel. C. Leave Bus 24-1 de-energized while the Maintena.--e Department repairs the Diesel Generator. D. No action required since no ECCS signal is prese,-: (the Diesel Generator is NOT supposed to auto start) Explanation: ANSWER:

b.

Attempt to manually start the #2 Diesel Generato -rorn the 902-8 panel. Explanation: With an ECCS signal present and the EDG not starting the correct action is to manually start the diesel from the 902-8 panel.

Reference:

DOA 6600-01 Question Pedigree: SBank 26400S0011I Cog level: High Rev 1 03/28/02 1 o1 Exam Submittal

Dresden 2002 11,1 NIC F'xam Q4 Exam System K/A RO SRO 60 BOTtI 295003 AK2.03 3.7 3.9 Partial or 2_ :replete Loss of A.C. Power _6i Objective: 262L-S 1-12 Knowledý--: )C the interrelations between A.C. electrical distribution system PARTIAL _,R COMPLETE LOSS OF A.C. POWER z:5the following: Given the lowing: L =: 2 is at 40% power. SV-* electric plant is in a normal line up. T E-* Unit 3 EDG is OOS for repairs to the governor. A fault de'. -ops on Bus 23 causing it to denergize. As a resul: _ ý'this... A. Bus 2f.ill be picked up by Bus 27. B. Bus 2-" r 27 will be picked up by Bus 25. C. Bus w.ill stay tied to Bus 23-I and be energized when the EDG starts and closes on Bus 23-1. D. Bus 2S will be load shed from Bus 23-1 and will have to be reclosed on Bus 23-1 after the EDG starts and closes ni Bus 23-1. Explanatio:. ANSWER

c.

Bus 2S will stay tied to Bus 23-1 and be energized when the EDG starts and closes on Bus 23-1. Explanati o:.: Bus 25 wfi be picked up by Bus 26. "a" is incorrect. Bus 23 was denergized so "b" is what would happen if Bus 24 los: Dower. Bus 28 does not load shed so "d" is incorrect. Reference0 2DM 262001 Question Pedigree: NEW Cog level: '-igah Rev I 03/28/02 1 of I Exam Submittai

Dresden 2002 l[f NRC Exam QI/ Exam System K/A RO SRO 61 BOTI-t 295004 AK3.02 2.9 3.3 Partial or Complete Loss of D.C. Power 70Obective: 263S-1,2-6 Knowledge of the reasons for the following [Ground isolation/fault determination responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: During performance of DOP 6900-06 125 VDC GROUND DETECTION - UNIT 2, the following ground detection meter indications are observed: the meter reads +40 volts with no buttons pushed the negative button is pushed and the meter goes to -60 volts the positive button is pushed and the meter goes to +100 volts The NLO then opens the U2 125 VDC TURB. BLDG. RESERVE BUS 2B-2 breaker on RESERVE BUS 2B. There is a known ground of+10 volts on RESERVE BUS 2B-2. How will Unit 2 ground detection indication respond and what is the reason for the response? A. All Unit 2 ground detection is lost because of the opened breaker. B. Unit 2 grounds stay the same because the deenergized bus is powered from Unit 3. C. The Unit 2 ground detector goes to +30 with no buttons pushed because of the known ground. D. The Unit 2 ground detector goes to +50 with no buttons pushed because of the known ground. Explanation: ANSWER:

b. Unit 2 grounds stay the same because the opened breaker is powered from Unit 3.

Explanation: The U2 125 VDC reserve turbine building loads are powered from U3. When the breaker in the question is opened it has no effect on the u2 125 VDC grounds.

Reference:

Question Pedigree: SDM 263002 and DOP 6900-06 New Cog level: High

Rev 2

03/28/02 1 of]1 Exam Submittal

Dresden 2002 IIF NRC Exam Exam BOTI i System 295005 K/A AA 1.04 RO 2.7 Main Turbine Generator Trip Objective: 245011 K )17 Ability to operate and/or monitor the following Main generator controls as they apply to MAIN TURBINE GENERATOR TRIP: A manual scram occurred on Unit 2 the following conditions are noted: The Main Turbine Stop Valves, Control Valves and Intercept Valves are c': _.ed. MWe on the 902-5 panel indicates -18 MWe There are no alarms up on the 923-2 panel Two minutes later conditions are the same. Based on these conditions the NSO will... A. start the Emergency Bearing Oil Pump. B. open the Main Generator OCB's. C. open the Turbine Vacuum Breaker. D. depress the Main Turbine trip pushbutton. 03/28/02 Exam Submittal I of I Q# 62 SRO

2.8 Explanation

ANSWER:

b. open the Main Generator OCB's.

Explanation: DGP 2-3 directs the NSO to open the OCB's if they have less than or equal to 0 MV-. on the 902-5 panel, with the turbine tripped over 90 seconds ago.

Reference:

DGP 2-3 Question Pedigree: Modified 03000S0421 Cog level: High Rev 1 I

Dresden 2002 1I,T NRC Vxaam 63 Exanm BOTH System "295005 Main Turbine Generator Trip Knowledge of the operational im: -:,tions of the following concepts as they appl\\ : ',[AIN TURBINE GENERATOR TRIP: The fuel is protected from damaz -:-rina a Main 03/28/02 I of I Exam Submittal K/A AKI 02 RO 3.2 SRO 3.6

Dresden 2002 ILT NRC Exam Q11 Exam System K/A RO SRO 64 BOaI 295031 EA2.01 4.6 4.6 Reactor Low Water Level Objective: 216L-S 1-3 Ability to determine and/or interpret the Reactor Water Level following as they apply to REACTOR LOW WATER LEVEL: Unit 3 is shutdown with the following conditions: No recirc pumps are running. Drywell temperature is 1 15'F. RPV pressure is 0 psig SDC pumps are secured. Which of the following is the lowest usable level indication available at the 903-5 panel to the NSO? A. -39 inches B. -51 inches C. -60 inches D. -295 inches Explanation: ANSWER

d. -295 inches.

Explanation: At rated conditions -295 is the lowest level that is available to the NSO at the 5 panel. Give the students a copy of DEOP 100 with the entry conditions blanked out. Refer ence SDM 216000 and Figures A, B and C of the Question Pedigree: DEOPs New Cog level: High Rev 2 03/28/02 I of I Exam Suom ittaI

Dresden 2002 IL" NRC Exam Q# 65 Exam BOTH System 295008 High Reactor Water Level Objective: 259-S 1-9 Knowledge of the reasons for the following Reactor feed pum~p trip: Plant-Specific responses as they apply to HIGH REACTOR WATER LEVEL: The feed pumps trip on high reactor level to prevent.... A. jet pump damage due to steam carryunder. B. HPCI turbine damage due to moisture carryover. C. feed pump damage due to cavitation and/or runout. D. main turbine damage due to moisture carryover. 03/28/02 I of I Exam Submittal K/A AK3.04 RO SRO

3.5 Explanation

ANSWER D Main turbine damage due to moisture carryover. Explanation: With high level in the vessel there will not be carryunder. The HPCI turbine is protected by a trip at 46 inches. The RFP trip occurs at 53 inches. The RFP are protected from cavitation and/or runout by a low suction pressure trip.

Reference:

SDM259L-S1, SDM223004, DAN 902(3)- Question Pedigree: 3-6 F-7 and 902(3)-3 A-9 Modified 25600S0051 Cog level: Memory Rev 1

Dresden 2002 ILT NRC. Exam SRO Exam Submittal 66 System 295009 AA-RO 3.6 SRO 3.7 Low Reactor Water Level i c yectiv7: 2591c-S-12 Ability to determine and/or interpret the Steare flur -e-ddflow mismatch following as they apply to LOW REACTOR WATER LEVEL: Unit 3 is a rated conditions when a transient occurs resulting following: C Total Steam flow is 9.8 MCbm/hr D Total Feed flow is 8.7 Mlbm/hr

• Reactor Water Level is currently at 27 inches.

What is the expected trend for Reactor Water Level and what ý-'-,edure will be required to be entered if operator actions to correct the problem are UNSUCESSFUL? LEVEL PROCEDURE A. Trending up DEO.- )0, RPV Control B. Trending up DOA,)-,Transient Level Control C. Trending down DEO7 '00, RPV Control D. Trending down DOA,0-,Transient Level Control Explanation: ANSWER:

c. Trending down, DEOP 100, RPV Control Explanation:

steamflow greater than feedflow would cause level to go down In entry into DEOP 100-1 RPV control would be needed when level reaches 8 inches.

Reference:

Quer.:on Pedigree: DOA 600-1 and DEOP 100 Ne\\ý Cog level: High Rev I 03/28/02 I of I

Dresden 2002 I[lT NRC Exam Q Exam System K/A R0 SRO 67 SRO 295010 AA2.06 3.6 3.6 HIgh Drywell P-sure Objective: 29501 LP004 Ability to deter---ie and/or interpret the Drywell temperature following as the. -apply to HIGH DRYWELL PRESSURE: Following a sta-: :. blackout event tile STA reports the following parameters to the Unit Supervisor. RPV le.-* -35 inches. drywell :c-nperature of 325'F drywell --essure of 6 psig Which of the fo" --wing action should be taken and what is the reason for that action? ACTION REASON A. Spray the D7-, well Convection cooling of the Drywell is needed to prevent over pressure condition in the drywell. B. Spray the Dr.,well Evaporative cooling of the Drywell is needed to prevent over pressure condition in the drywell. C. Blowdown Evaporative cooling would result in drywell pressure reducing to less than 2 psig and possible implosion of the Drywell. D. Blowdown Convection cooling would result in drywell pressure reducing to less than 2 psig and possible implosion of the Drywell. Explanation: ANSWER:

c.

Blow down Evaporative cooling resulting in drywell pressure reducing to less than 2 psig and possible implosion of the D-ywell. Explanation Spraying the Dry'v, ell with these conditions would rapidly lower pressure in the drywell with the potential for implosion. The mechanism the this would be caused by is evaopriitive cooling not convection cooling.

Reference:

DEOP 200-1 and BWROG EPGs Question Pedigree: New Cog level: High Rev I 03/28/02 1 of I Exam Submittal

Dresden 2002 ILT NRC lxam Exam BOTII System 295013 IK/A AA2.01 RO 3.8 SRO 4.0 F-ligh Suppression Pool Temperature Obj'ecti~ve:22 33L-19 Ability to deter-ine and/orr interpret the Suppression pool temperature following as they apply to HIGH SUPPRESSION POOL TEMPERATURE: Unit 2 is at rated conditions with the 902-36 back-panel recorder TIRS 2-1640-200A, ýTORUS TEMP M, DIV I OOS due to a failed power supply and all appropriate Technical Specifications have been entered. T[RS 2-1640-200B currently indicates the following:I Point I 1 12°FF Point 5 850F/ Point 2 95 0F Point 6 850F Point 3 90OF Point 7 870F Point 4 850F Point 8 907F Explanation: ANSWER

a.

No actions are required at this time. Explanation: Bulk water temperature is the average of points I through 8. Average water temperature would be 91.125°F. No action is required. The requirements of ITS section 3.6.2.1 specifically state that the temperature is average temperature, not the highest of any one area.

Reference:

295L-$2 and ITS 3.6.2.1 Question Pedigree: Last years NRC Exam question 71 Cog level: High Rev 03/28/02 I of I Exam Submittal Q8 68

Dresden 2002 lIT NRC Exam Q8 69 Exam BOTu I System 295014 K/A AK2.01 RO 3.9 SRO 4.1 D. Reactor power would decrease and stabilize due to the change in void fraction. Explanation: ANSWER

a.

The reactor would scram due to high flux. Explanation: Large increases in reactor pressure at the above conditions would result in a reactor scram c_- to high flux.

Reference:

DGP 01-01 page 19 Question Pedigree: Last years NRC Exam #72 Cog level: High ?e I of I 03/28/02 Exam Submittal Inadvertent Reactivity Addition Objective: 24501 LK021 Knowledge of the interrelations between RPS INADVERTENT REACTIVITY ADDITION and the following: A startup is in progress on Unit 3 with the following conditions: Reactor pressure is 170 psig. One bypass valve is full open. Control rods are being withdrawn to achieve two bypass valves open. IRMs are between 30 and 70 on range 8 Which of the following would be expected to occur if all bypass valves were to fail closed no operator action? A. The reactor would scram due to high flux. B. The reactor would scram due to high pressure. C. Reactor power would increase and stabilize due to the change in void fraction.

I)- den 2002 IL TNRC Ixam Exam BOTI I )-5 14 Inadvertent Reactivity -Addition Knowledge of the reasons for the follo\\x 2 responses as they apply to INADVERTEP-T REACTIVITY ADDITION: Exam Submittal 70 K/A AK3.02 R, -3.7 S RO 3.7 03/28/02 I of I

Dresden 2002 IlT NRC Exam Exam SRO System 295015 K/A 2.4.30 RO 2.2 SRO 3.6 Incomplete SCRAM Objective: 29900LK026 Emergency Procedures and Plan Knowledge of which events related to system operations/status should be reported to outside agencies. Unit 3 was at rated conditions when the NSO reports: Immediate actions for responding to a reactor water level of -4 inches are complete. All rods are in. The A and B RPS solenoid group lights are lit. Which of the following lists or identifies the HIGHEST level of notification that is required to be made for this condition? A. Illinois EMA per EP-AA-1 14 Notifications B. Plant Manager per OP-AA-106-101 Significant Event Reporting C. Illinois EMA AND Grundy County Sheriff per EP-AA-114 Notifications D. Plant Manager AND Site Vice President per OP-AA-106-101 Significant Event Reporting Explanation: ANSWER a Illinois EMA per EP-AA-1 14 Notifications Give the students page DR 3-3 of the EALs Explanation: The key to the questions is the If rods go in when ARI is initiated manually the event is classified as an Alert not a Site Emergency. OP-AA-106-101 Significant Event Reporting does not apply since this is an EP classification (section 4. 1.1 )

Reference:

Question Pedigree: EALs MS3 and MA3, EP-AA-114, OP-AA-106-101 New Cog level: High Rev 2 03/28/02 Exam Suo-

tal I of I Q#

71

Drcsden 2002 1" F NRC Exam Q#! Exam System K/A RO SRO 72 BOTH 295015 AK 1.04 3.8 3.8 Incomplete SCRAMN Objective: 29501 LK03 I Knowledge ofthee operational implications of the Reactor pressure: Plant-Specific following concepts as they apply to INCOMPLETE SCRAM: An automatic scram occurred on Unit 3 Control rods did not fully insert and reactor power decreased to 10% Containment parameters will require an emergency depressurization within fifteen minutes if trends are not changed. Opening the bypass valves now to rapidly reduce reactor pressure should.... A. be performed to allow for the reduction of reactor power. B. be performed to anticipate an emergency depressurization. C. NOT be performed since the pressure reduction will add significant positive reactivity. D. NOT be performed since the pressure reduction will result in removal of boron from the RPV. Explanation: ANSWER

c. NOT be performed since the pressure reduction will add significant positive reactivity.

Explanation: With the reactor still at power, the rapid depressurization will add significant positive reactivity to the core complicating the power actions underway. It is for this reason that an emergency depressurization is only performed if the conditions that require it are actually met.

Reference:

295L-S I Question Pedigree: Last years NRC exam #99 Cog level: Memory Rev I 03/28/02 1 of I Exam Submittal

Dresden 2002 ILT NRC Fxam System 295016 K/A 2.1.32 Control Roomn Abandonment Conduct of Operations Objec ,-: 29501RK001 Ability to explai:- :d apply system limits and precautions. During the performance of DSSP 0100-CR "Control Room Evacuati,- it is reported that the Isolation Condenser Make-up Pumps will not start. According to the UFSAR this is a concern because (2) A. damage will occur to the Isolation Condenser tubes B. damage will occur to the Isolation Condenser tubes ___(1)____ ._nd this is prevented by 2 -I isolating the.:olation Condenser from the reactor vessel adding make _p to the Isolation Condenser from another sour:. C. inventory in the vessel will be lost isolating the '-olation Condenser from the reactor vessel D. inventory in the vessel will be lost adding make _p to the Isolation Condenser from another sourc.- Explanation: ANSWER D. inventory in the vessel will be lost AND adding make up to the Isc'tion Condenser from another source Explanation: There is a NOTE in the Control room abandonment procedure that sa%_ make up to the isolation condenser must be initiated within 20 minutes. The UFSAR states with out the Is_- condenser to lower pressure inventory will be lost when the ERVs lift to control pressure.

Reference:

DSSP 100-CR and UFSAR section 5.4.6.3 Question P-' igree: New Rev 1 I of I Exam Submittal Q4 73 Exam SRO ' RO 3.4 SRO 3.8 1.a I I. 1*511 03/28/02 I I

Dresden 20021 Ili NRC [-xam QL ExL-System K/A R.O SRO 74 B0_- 295016 AA 1.04 3.1 3.2 Control Room Abando- -nt Objective: 262L-S4-12 Ability to operate and "onitor the following A.C. electrical distribution as they apply to CONT- -)L ROOM ABANDONMENT: DSSP 0100-CR "Contr: Room Evacuation" is in progress. How is the Bus 29-28 - reaker closed? A. Depress the manua :oepushbutton oil the front of the breaker. B. Plug in the local pL -' utton control station and depress the close button. C. Place the two hooks -- ithe operating handle in the lower portion of the cubicle and push down on the operating toon. D. Place the ratchet typ m aintenance tool on the shaft that protrudes from the breaker and operate the handle until the breaker clc,ss. Explanation: ANSWER:

b. Place the ratchet type -.aintenance tool on the shaft that protrudes from the breaker and operate the handle until the breaker closes.

Explanation: The Bus 29-28 TIE reqL -es the ratchet type tool to close the breaker.

Reference:

DSSP 100-CF Question Pedigree: New Cog level: Memory Rev I 03/28/02 I of I Exam Submittal

l)resden 2002 II[T NRC I'xam 75 Exam BOTIH System 295017 K/A A K2.04 RO 3.1 SRO 3.3 [-igh Off-Site Release Rate Objective: 288L-S I-05 Knowledge of the interrelations between HIGH Plant ventilation systems OFF-SITE RELEASE RATE and the following: Which ONE of the following conditions will cause a Reactor Building Ventilation Isolation? A. A 10 R/hr radiation level in the drywell. B. A 10 mR/hr radiation level on the refuel floor. C. An upscale trip on one Reactor Building ventilation radiation monitor. D. A downscale trip on one Reactor Building ventilation radiation monitor. Explanation: ANSWER:

c. An upscale trip on one Reactor Building exhaust plenum monitor.

Explanation: It take 100 mrem on the refuel floor, or 30 R in the drywell, or two downscale trips to cause an isolation.

Reference:

DOA 5750-01 Question Pedigree: Modified 288801 SO 11 Cog level: Memory Rev 2 03/28/02 -xam Submittal I of I

Dresden 2002 ILT NRC Exam Q/1 Exam System K/A RO SR 76 RO 295017 A K3.03 3.3 4.5 High Off-Site Release Rate Objective: NONE Knowledge of the reasons for the following Implementation of site emergency plan responses as they apply to HIGH OFF-SITE RELEASE RATE: The reason for the implementation of the General Emergency plan is to... A. minimize the damage done to plant equipment. B. allow personnel to exceed the normal exposure limits. C. allow the NRC to be involved in decisions made at the plant. D. minimize or prevent exposure above federal limits to offsite personnel. Ezxplanation: ANSWER:

d. minimize or prevent exposure above federal limits to offisite, personnel.

Explanation: The heath and safety of the public is the reason for GSEP

Reference:

Question Pedigree: EP-AA-113 and EP-AA-1 11 New Cog level: Memory Rev I 03/28/02 I of I Exam Submittal

Dresden 200-NRC Exam Q4 Exam System K/A R) SRO 77 RO 295018

2. 1. 14 2.5 3.3 Partial or Complete Loss of Component Cooline \\V--- -

Objective: 209L-SI-5 Conduct of Operations %..V-ledge of system status criteria which require the - -:-fication of plant personnel. Unit 3 is operating at rated conditions when the fo]!, ng occurs: The 3A RBCCW pump trips. The 3B RBCCW pump is successfully stariý.- -nd all RBCCW system parameters return to normal. There is no indication of an electrical trip or --- ý breaker for the 3A RBCCW pump. What is the lowest level of authority that must autho.-:2 a restart of the 3A RBCCW pump? A. Only the Unit Supervisor B. The Shift Manager and Engineering C. The Shift Manager and Electrical Maintenance D. The Unit Supervisor and Electrical Maintenance Explanation: ANSWER

a. Only the Unit Supervisor Explanation:

With no electric fault indicated only the US authoriz.-.n is needed to attempt to restart the pump,

Reference:

Question Pedigree: DOA 6500-10 New Cog level: Memory Rev 2 03/28/02 1 o5 Exam Submittal

Dresden 2002 ILT NRC Exam Q1-78 Exami SRO System 295019 K/A AA2.02 RO 3.6 Partial :r Complete Loss of Instrument Air 1 Objective: 20 1..-S 1-7 Abilit\\ :. determine and/or interpret tile Status of safety-related instrument air system a follo\\* :.-. as they apply to PARTIAL OR AK2. I -AK2.19) CONMPETE LOSS OF INSTRUMENT AIR: Jnstruirent Air on Unit 3 has been lost to the Scram Discharge Volume (SDV) vent and drain valves. Unit, 7-mains at 100% power. It is ex\\-cted that the SDV vent and drain valves will fail.... A. CL OhSED and be INOPERABLE since the SDV would be isolated firom the scram outlet header! SRO 3.7 03/28/02 I of 1 Exam Submittal

Dresden 2002 IIT NRC Exam System 295019 K A A1K2.17 Partial or Complete Loss of Instrument Air Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: What effect will the loss of Instrument Air have on th Objective: 206L-S 1-12 [igh pressure coolant injection: Plant-Specific e HPCI system? A. HPCI Turbine Exhaust Pot Bypass Valve (2301-28) will fail closed. B. HPCI Steam Line Drain Trap Bypass Valve (2301-31) will fail open. C. Turbine Steam Supply Line Drain Valves (2301-29, -30) will fail open. D. Turbine Stop Valve Above Seat Drain Valves (2301-64, -65) will fail closed. 03/28/02 I of I Exam Submittal Q9 79 BOTH RO 2.7 SRO Explanation: ANSWER:

d.

Turbine Stop Valve Above Seat Drain Valves (2301-64, -65) will fail closed. Explanation: All of the valves listed in this question fail to the HPCI operating position on a loss of Instrument Air. The 2301-28 fails open, the 2301-31 fails closed, and the 2301-29, -30 fail closed. The 2301-64,-65) will fail as stated in the closed position.

Reference:

SDM206000 Question Pedigree: New Cog level: Memory Rev

Drcsdcen 2002 ILT NRC Exam System 295020 IK/A AA2.01 Inadvertent Containment Isolation Ability to determine and/or interpret the following as they apply to INADVERTENT CONTAINMENT ISOLATION: I -- Objective: 239S-L ontainment pressutn Unit 2 is at rated conditions and has been for the last 300 days. An inadvertent Group I isolation signal is received on 'ligh Main Steam Line Fk What is the FIRST thing this causes? A. EDGs to start. B. HPCI to start. C. Drywell pressure to go up. D. Torus temperature to go down. Explanation: ANSWER

c. drywell pressure to go up.

Explanation: With EPU on Unit 2 the ERV will open to control pressure. This will cause drywe' -ressure to go up. The EDGs will not recive a start signal unless drywell pressure reaches 2 psig. HPCI w not start till level drops to -59 inches and torus temperature will go up not down.

Reference:

299L-S5 Question Pedigree: New Cog level: Hi-gh 03/28/02 Rev 2 1 of I Exam Submittal Q// 80 Exam RO RO 3.6 SRO 3.7 I

Dresden 2002 ILI" NRC E'xam System 295022 K,\\ AA2.01 Loss of CRD Pumps I Objective: 201 L-S 1-01 Ability to determine and/or inte]T--z the following as they apply to LOSS F CRD PUMPS: Accumulator pressure Unit 3 was at rated conditions xvi-.- ALL equipment available when the following events occurred: The running CRD pump -T-?ped. Two peripheral control ro: 'ACCUMULATOR TROUBLE" alarms are received, The following additional informna:: a, is provided: The two controls rods are - notch 48 Accumulator pressure for :.e alarming accumulators is 925 psig. Which of the following describes e NEXT action that should be performed and the reason for the action? ACTION REASON A. Scram the reactor To prove the ability of the CRD system to scram the reactor without the reliance on the CRD drive water. B. Scram the reactor To ensure Shutdown Margin requirements are met should the controls rods associated with the failed accumulators fail to insert. C. Start the standby CRD pump _--d verify To prove the ability of the CRD system to supply drive charging water header pressure is at least water pressure to insert the control rods without the 940 psig accumulators. D. Start the standby CRD pump ::id verify To prevent damage to the control rod drive mechanisms charging water header pressure is at least due to overheating. 940 psig 03/28/02 I of I Exam Submittal 81 Exam SRO RO 3.5 SRO

3.6 Explanation

ASWER

c. Start the standby CRD pump an: verify charging water header pressure is at least 940 psig (ACTION) and To prove the ability of the CRD s\\.:em to supply drive water pressure to insert the control rods without the accumulators.(REASON)

Explanation Immediate action of DOA 300-01 ; to start standby CRD pump. With accumulators inoperable, operators must prove the ability to insert control r-ds.

Reference:

DOA 300-01 Question Pedigree: Modified Last years NRC exam #83 Cog level: High Rev 2

D)resden 2002 II.'1 NRC E'xam 82 BOlTH 03/28/02 Exa Dm jitt a I System 295023 K/A 2.1.14 RO 2.5 SRO 3..3 Refueling Accidents Obective: 23400LK001 Conduct of Operations Knowledge of system status criteria which require the notification of plant personnel. Unit 2 is in a refueling outage when the refuel floor radiation alarm sounds. The AUX NSO reports the Refuel Floor ARM indicates 110 torern/hr. Which of the following actions must be taken? A. Evacuate the Refuel Floor only. B. Evacuate the Refuel Floor and the Reactor Building. C. Notify the Fuel Handling Supervisor that the alarm is erroneous. D. Notify the Fuel Handling Supervisor that the alarm is valid and work may continue with caution. Explanation: ANSWER

a.

Evacuate the Refuel Floor only. Explanation: The action required by DFP 0850-03 are to evacuate the refuel floor when the ARM indicates greater than 100 mrem.

Reference:

DFP 0850-03 Question Pedigree: New Cog level: High Rev I I ofI

Dresden 2002 IT NRC Exam QI/ Exam System K'A RO SRO 83 BOTH 295023 AA 1.03 3.3 3.6 Refueling Accidents Objective: 23400LK01 Ability to operate and/or monitor the following Fuel handling equipment as they apply to REFUELING ACCIDENTS: The reactor building overhead crane and the refueling bridge crane are being used to move equipment during refueling outage when radiation levels reach 40 inrem / hr on the refuel floor. What are the consequences of the radiation level? A. Standby Gas Treatment will auto start. B. The Reactor Building Ventilation system will isolate. C. The refueling bridge crane will be prevented from raising fuel. D. The reactor building overhead crane hoist raise function is inhibited. Explanation: ANSWER

d. The reactor building overhead crane hoist raise function is inhibited.

Explanation: Standby Gas Treatment and Reactor Building Ventilation auto actions occur at 100tmrem /hr. High radiation is not a refueling bridge crane interlock.

Reference:

Question Pedigree: DFP 850-03 New Cog level: High Rev I 03/28/02 1 of I Exam Submittal

Dresden 2002 I [,' Exam OilExam System A' RO SRO 84 BO'I I 295024 EL.01 4.l 4.2 i gh Drywell Pressure Objective: 223S L 1-12 Knowledge of the operational implications of the Dr\\\\. ntegrity: Plant-Specific following concepts as they apply to HIGH DRYWELL PRESSURE: Unit 2 is at rated conditions when the following occurs: A design basis Loss of Coolant Accident (LOCA). A Drywel to Torus Vacuum Breaker fails open. This will FIRST result in drywell pressure... A. equalizing with Reactor Building pressure. B. exceeding the design pressure of the containment. C. dropping rapidly since the cooling effectiveness of the 1: -us has been greatly improved. D. staying below the design pressure since there are a num'._r of redundant vacuum breakers installed. Explanation: ANSWER:

b.

exceeding the design pressure of the containment. EXPLANATION: A failed open vacuum breaker will allow steam to bypass the -ippression pool thereby causing the pressure spike to exceed design pressure.

Reference:

SDM223001 Qu~stion Pedigree: Ba7'(i 22301S0311 Cog level: High ev2 03/28/02 I of I Exam Submittal

Dresden 2002 IL T NRC Exam Q11Exam System K/A RO SRO 85 SRO 295025 2.4.4 4.0 4.3 High Reactor 7---ssure Objective: c Emergency P::.'-dures and Plan Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency ard abnormal operating procedures. The followinL - :nditions exist on Unit 3: RPV "t--1 is 26 inches and rising. RPV F--sRure is 1070 and steady. All roL- -re in. EHC

-,sSure is 0 psig.

The N --,Is are OPEN and the bypass valves are CLOSED. With these ind.: cations the operating team should FIRST en'ter... A. DEOP 10(--'. RPV Control, and restore level using HPCI. B. DEOP I0*c-- *. RPV Control, and initiate the Isolation Condenser, C. DOA 600-Transient Level Control, and restore level by starting the standby Condensate/condensate booster pu-.:s. D. DOA 565sc--. Pressure Regulator Failure, and reduce reactor pressure with pressure set. Explanation: Answer:

b. DEOP 100-1 RPV Control, and initiate the Isolation Condenser.

Explanation: Meets SRO on'. criteria number 5. For the conditic -; given, the action that need to be taken first are restoring pressure in accordance with DEOP 100-1.

Reference:

DEC ? 100-I Question Pedigree: New Cog level: High Rev I I of I 03/28/02 Exam Submittal

Drcs'dcn 2002 IILT NRC Exam Svstelmi 295025 K/A ElK3.04 High Reactor Pressure Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE: Objective: 207L-SI -I Isolation condenser initiation: Plant-Specific Given the following conditions on Unit 3: 0 0 0 The plant had been operating at 100% for 6 months. A Group I isolation occurred ten minutes ago. All AC power has been lost to Unit 3. Which of the following systems is designed to provide reactor pressure control/cooling under these conditions? A. HPCI B. Isolation Condenser C. Main steam line drain valves D. Automatic Depressurization System I of I Exam Submittal Q/1 86 Exam BOTI I RO 4.5 SRO

4.7 Explanation

ANSWER

b. Isolation Condenser Only the IC is designed specifically for the conditions described in the stem. HPCI's purpose is medium LOCA, ADS is a backup to HPCI, and the MSL drains are not available (due to loss of A/C power).

Reference:

Question Pedigree: SDM 207000 Bank 20700S0401 Cog level: Memory Rev I 03/28/02

Dresden 2002 [L'U NRC Exam Q/ Exam System K/A RO SRO 87 IOTH 295028 EK 1.02 2.9 3.1 High Drywell Temperature Objective: 29011LK02 Knowledge of the operational implications of the Equipment environmental qualitlca following concepts as they apply to HIGH DRYWELL TEMPERATURE: Unit 3 is experiencing a LOCA. The following conditions exist: reactor is shutdown drywell pressure is 10 psig drywell temperature is 350'F (point 9) reactor pressure is 75 psig reactor water level is -45 inches reactor building temperature is 105'F Fuel Zone level indication is OOS Which ONE of the following is the reason that RPV water level indication may NOT be :--able? A. Drywell pressure is excessive. B. Drywell temperature is excessive. C. Reactor Building temperature is excessive. D. RPV level is below minimum usable indicating levels. Explanation: ANSWER:

b. Drywell temperature is excessive.

Reference:

DEOP 100 Figure A and B Modified 29502S0881 Cog level: High Rev 2 03/28/02 1 of I Exam Submittal

-den 2002 ILT NRC Fxam Q Exam ýtern K/A RO SRO 88 RO -029 FA2.02 3.5 3.6 High Suppression Pool Water 2evel Objective: 29502LK004 Ability to determine and/or interpret :-:Reactor pressure following as they apply to HIGH SUPPRESSION POOL WATER LE', = Conditions are as follow on Unit 2 , Torus water level is 19 feet. All efforts to lower Torus le\\ z -ave failed. Which of the following events could -.: -ardize containment integrity? A. Core Spray initiation and injectio B. ERVs cycling on high reactor pre ::re. C. HPCI initiation with suction froir ;e torus. D. Lining up the LPCI system for to: cooling. Explanation: ANSWER:

b.

ERVs cycling on high reactor pre:.:-:re. Explanation: If torus level can not be held < 18.5 fe-. all outside injection sources must be stopped. If the relief valves were to lift damage to the containment coul: :ccur.

Reference:

295L-$2 Question Pedigree: New Cog level: High Rev1 03/28/02 I of I Exam Submittal

Dresden 2002 ILT NRC Exam Q#P Exam System K/A RO SRO 89 SRO 295029 EA2.03 3.4 3.5 1-igh Suppression Pool Water Level Objective: 223L-S1-07 Ability to determine and/or interpret the Drywell/containment water level following as they apply to HIGH SUPPRESSION POOL WATER LEVEL: Unit 3 was at rated conditions when a LOCA occurred. The Aux NSO makes the following report to the Unit Supervisor, " Torus level rose to 20 feet immediately after the LOCA occurred then returned to a level of 15 feet 3 to 5 seconds later." This report indicates that... A. DEOP 400-2 should be entered due to high Torus level. B. the Unit is experiencing a LOCA outside the design bases. C. an NLO must be dispatched to locally determine Torus level. D. "Pool Swell" has occurred as described in the design bases. Explanation: ANSWER

d. "Pool Swell" has occurred as described in the design bases.

The UFSAR explains this phenomenon and the design basis accounts for its occurrence. Locally verifying level with the sightglass is not required. The DEOP entry is not required unless level cannot be held below 18.5 feet. This meets SRO criteria # 1

Reference:

UFSAR section 6.2.1.3.4.1.1 Question Pedigree: New Cog level: High Rev 1 03/28/02 I of] Exam S-.mmittal

Dresden 2002 I.T NRC lFxam Q/ Exam System K/A RO SRO 90 SRO 295030 EA2.04 3.5 3.7 Low Suppression Pool Water Level Objective: 29502LK002 Ability to determine and/or interpret the Drywell /suppression chamber differential pressure: Mark following as they apply to LOW l&Jl SUPPRESSION POOL WATER LEVEL: The following conditions exist on Unit 3. Torus level is -4.4 inches Drywell to Torus dP is 1.6 psig Annunciator "TORUS NARROW RANGE WTR LVL LO" 902-4 C-23 is alarming. Which of the following procedures must be entered? A. DEOP 100 "RPV Control" B. DEOP 200-I "Primary Containment Control" C. DOP 1600-1 "Normal Pressure Control of the Drywell and Torus" to vent the Torus D. DOS 1600-2 "Torus Level Verification using Local Sightglass" to validate the control room indicator. Explanation: ANSWER

b. DEOP 200-1 "Primary Containment Control" Explanation:

Meets SRO only criteria #5. This is not a listed entry condition on DEOP 200-1, however in DOP 1600-2 Torus Water Level Control states that if dP and level are not within the curves on Attachment A DEOP 200-1 must be entered. Give the students a copy of DOP 1600-2

Reference:

Question Pedigree: DEOP 200-1 and DOP 1600-2 New Cog level: High Rev 2 03/28/02 1 of I Exam Submittal

Dresden 2002 ILT NRC I Q# Exam System K A RO SRO 91 BOTI f 295031 FK2. ;- 4.1 4.1 Reactor Low Water Level ( ctive 259S L212 Knowledge of the interrelations between Reactor wa:, - - el control REACTOR LOW WATER LEVEL and the following: Unit 2 is at rated conditions when the following occurs: a A transient occurs that causes the following RPV leve! -. -ation: Narrow Range A indicates -2 inches. 0 Narrow Range B indicates +1 inches. a Medium Range A indicates -I inches. A Reactor Scram does NOT occur. What is the response of the FWLC system? A. Stays in Master Auto and attempts to restore level. B. Sets RPV level setpoint to +5 inches immediately. C. Ramps RPV level setpoint to +5 inches at 10 inches/min. D. Enters Master Manual and the operator must restore level. Explanation: ANSWER:

b. Sets RPV level setpoint to +5 inches immediately.

Explanation: Sets RPV level to +5 inches if FWLCS is in Master Auto and the ' owing occur

1. Reactor Scram OR

2.

Two of the following occur: "* Narrow Range A indicates <0 inches. "* Narrow Range B indicates <0 inches. "* Medium Range A indicates <0 inches.

Reference:

Questi,:- Pedigree: DOA 600-I and SDM 259002 New Cog level: High Rev 1 03/28/02 1 of I Exam Submittal

Dresden 2002 II F NRC Fxam System 295032 K/A 2.4.49 High Secondary C ý -tainment Area Temperature Emergency ProceZ_-es and Plan I Objective: 29502LP016 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. Unit 2 is at rated c: Jitions when the following occurs: Annuncia: -. "CHANNEL A MN STM TUNN TEMP I-I" 902-5 D-9 alarms. The NSO -r-orts the Shutdown Cooling Pump Room temperature is 190'F and rising slowly. The Unit Supervis: - should direct the NSO to... A. increase TBCC',V flow to the X-Area coolers. B. manually scra-T the reactor and perform a blowdown. C. manually scra-the reactor and shut the MSIVs. D. secure Reactor Building ventilation and start SBGT. 03/28/02 I of I Exam Submittal Q9 92 IRO RO 4.0 SRO

4.0 Explanation

ANSWER: c manually scram -.e reactor and shut the MSIVs. Explanation: These are the correc: action per the DEOP 300-1 flow chart. TBCCW does not cool the X area coolers, a blowdown is not recuired until TWO areas are above Max Safe and starting SBGT would not cool the area down. Give the students a copy of DEOP 300-1 with the entry conditions blanked out.

Reference:

DEOP 2 I Question Pedigree: New Cog level: High Rev I

Dresden 2002 IT NRC Exam Q# Exam System K/A RO SRO 93 BOTI! 295032 EA 1.03 3.7 3.7 IHigh Secondary Containment Area Temperature Objective: 288S-1,2-5 Ability to operate and/or monitor the following Secondary containment ventilation as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE: After a trip of Reactor Building ventilation it has been determined that Reactor Building ventilation needs to be restarted per DEOP 300-1, Secondary Containment Control, to lower Reactor Building temperature. What must be done to restart Reactor Building ventilation? A. Two vent fans must be started first. B. Two exhaust fans must be started first. C. Install jumpers to bypass High Reactor Building Temperature isolation. D. Fan control switch must be held in CLOSE for a minimum of five seconds. Explanation: ANSWER:

d. Fan control switch must be held in CLOSE for a minimum of five seconds.

Explanation: There is no high RB temperature isolation. The fans are started one exhaust the one vent. The switch must be held for 5 seconds to allow air flow to develope.

Reference:

Question Pedigree: DEOP 500-2 New Cog level: Memory Rev I 03/28/02 I of ] Exam Suomittal

Dresden 2002 ILT NRC [\\xam Q/1 94 Exam BOTH System 295033 K/A EA 1.0 1 RO

3.9 Explanation

b. after the ARM BYPASS SWITCH for the ARM in alarm has been placed in the BYPASS pc,: :on.

Explanation: The toggle switches above the multi-point recorder allow for individual points to be bypassed.: -:e operator would be aware of other alarms.

Reference:

Question Pedigree: SDM 272001 Bank 27201 S0061 Cog level: Memory Re" 03/28/02 Exam Submittal High Secondary Containment Area Radiation Levels Objective: 2721,-S 1-6 Ability to operate and/or monitor the following Area radiation monitoring system as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS: Given the following conditions: The 902-3 A-I "RX BLDG RAD HI annunciator has alarmed. The ARM in alarm is still above its alarm setpoint. The annunciator will be able to be reset... A. ONLY when that ARM's RESET button is depressed on the 902-I panel Indicator and Tr:.- -'nit. B. after the ARM BYPASS SWITCH for the ARM in alarm has been placed in the BYPASS -,-ition. C. ONLY when that ARM's SILENCE button has been depressed locally at the auxiliary unit D. after acknowledging the 902-3 panel annunciator and then by depressing the 902-3 panel F --:-t pushbutton. I of I I

Dresdc' " 02 II,'F NRC Exam Fxam BOTI l System 2950;3 EK/A EK1.01 RO SRO 4.1 Secondary Containment Ventilation High Raý on Objective: 29502LK050 Knowledge of the operational implications of:1* Personnel protection following concepts as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: Unit 3 was at rated conditions when a transier: ccurred. An Isolation Condenser steam leak occur. -: and was isolat-1 Isolation Condenser area temperature is I -. 'F and is too high for personnel access. Valid Reactor Building Ventilation isolar as are present on each of the following parameters: Drywell Pressure Reactor Building Exhaust Radiation Reactor Water Level Restarting the Reactor Building Ventilation wcd allow safer access to the Isolation Condenser area... A. but is NOT allowed due to the Drywell Prz. ure isolation. B. but is NOT allowed due to the Reactor Bul 'ing Exhaust Radiation isolation. C. but is NOT allowed due to the Reactor WVa:ar Level isolation. D. and maybe performed after bypassing the.-olation signals. Explanation: ANSWER

b.

But is NOT allowed due to the Reactor Bu:'.ding Exhaust Radiation isolation. Explanation: Only the drywell and RPV water level isolatiors are allowed to be bypassed since they do not indicate a release hazard. Reactor building exhaust radiation aboN z the isolation setpoint would be indicated of a potential radioactive release problem and would not be al-wed to be bypassed. Question Pedigree: Last years NRC Exam #97 Cog level: Memory 03/28/02 Rev I I ofI Exam Submittal Q# 95 Rev I Rererence: 295L-S3

I)rCsdcn 2002 I1, L NRC Exam Exam System K/A RO SRO BOTH 295035 FK3.02 S 0 --: )ndary Containment Hi gh Differential Pressure Oei 288S-L 1-6 wledge ofthe reason s for--i t c v ton response -onses as they apply to SECONDARY S C NTAINMENT HIGH DIFFERENTIAL P;-_>SSURE: -., do the Reactor Building Ventilation supply fans trip on h iugh R t B i p s A. To prevent an auto initiation of SBGT. B. To prevent actuation of the Reactor Building blowout panels. C. To ensure that airflow is from high contamination to low contamination. D. To prevent damage to the Reactor Building Ventilation supply fans butterfly dampers. Ex-Ination: AN WER:

b. T, prevent actuation of the Reactor Building blowout panels.

Ex:.anation: Thm -ans trip at 2.2 inches H20, this protects the building from overpressure. If an overpressure condition was to d-velop the RB blowout panels would activate at 13 inches H20 to protect the secondary containment. Refe-rence: Question Pedigree: SDM 288001 and 223001 and LP 288S-L1 New Cog 'vel: Memory Rev 2 03/28 I of l Exam Submit7-*

Drcsden 2002 II.T NRC Exam Q!/ Exam System K/A RO SRO 97 BOT[1 295037 EA 1.04 4.5 4.5 SCRAM Condition Present and Reactor Power Above Objective: 211 L-S 1-06 APRM Downscale or Unknown Ability to operate and/or monitor the following SBLC as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: The following conditions exist: A transient has occurred on Unit 2. The Unit Supervisor has ordered both SBLC pumps started for injection into the reactor. After placing the INJECTION CONTROL switch to the SYS l&2 position, the operator observes the following indications: "* The SQUIB A and SQUIB B lights are lit. "* The Pump 1 and Pump 2 lights are lit. Which ONE of the following, is the proper course of action in order to attempt to get full initiation of BOTH SBLC subsystems? A. Dispatch an NLO to start the SBLC pumps locally. B. Dispatch an NLO to manually open the SQUIB valves. C. Position the INJECTION CONTROL switch to the SYS I position. D. Position the INJECTION CONTROL switch to the SYS 2&l position. Explanation: ANSWER:

d.

Position the INJECTION CONTROL switch to the SYS 2&l position. Explanation: There is no ability to manually open the squib; "b" is incorrect. Taking the switch to the SYS 1 position will not allow both subsystems to inject, "c" is incorrect The pumps are already running; "a" is incorrect. The reason the squib valve may not have fired might be due to a circuitry problem, taking the switch to the other dual pump position may successfully fire the squib; "d" is correct.

Reference:

SDM21 1000 Question Pedigree: Modified 21100S0171 Cog level: High. Rev I I of I 03/28/02 Exam Submitta8

Dresden 2002 ILT NRC Exam Q Exam System K/A

• O SRO 08 BOTfI 295037 EKI.02 4.3 SCRAM Condition Present and Reactor Power Above Objective:

APRM Downscale or Unknown Knowledge of the operational implications of the Reactor water level etY-:. on reactor power following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: When cloding the RPV during a Failure to Scram condition, why must i -oninto the RPV be increased slowly? A. To minimize the thermal shock the clad experiences. B. To prevent a large power transient that may cause core damage. C. To allow the operators time to ensure the Main Steam lines do not beco7-. flooded. D. To ensure the reactor does not pressurize uncontrollably when the reactc 7goes solid. Explanation: ANSWER:

b.

To prevent a large power transient that may cause core damage Explanation: Since the reactor may become critical with addition of cold feedwater the wa:er must be added slowly to prevent a large transient on the core.

Reference:

Question Pedigree 295L-$4 Bank 35900S008. Cog level: Memory Rev 1 03/28/02 I of I Exam Submittal

0 l)resden 2002 ILT NRC Fxam Q Ex am System K/A RO SRO 99 BOTH 295038 EA 1.03 3.7 3.9 high Off-Site Release Rate Objective: 272[-S2-8 Ability to operate and or nmc - -)r the following Process liquid radiation monitoring system as they apply to HIGH OFF-- -E RELEASE RATE: S Unit 2 is at rated conditions -n the "LIQUID PROCESSRA P 902-3 G-1 annunci, Where can tihe operator dete-- =e the actual RBCCW effluent monitor levels? A. 902-10 panel - control ro:-= backpanel B. 923-7 panel - SPING pa.r C. Rad Waste Control Roorz D. at the RBCCW expansior. -n-,k Explanation: ANSWER

a. 902-10 panel - control room :-,-.ckpanel Explanation:

The rad monitor for RBCCW _cated at the 902-10 panel.

Reference:

Question Pdigree: DAN 902(3)-3 G-I New Cog level: Memory Rev 2 03/28/02 1 of I Exam Submittal

Dresden 2002 lIT NRC Exam Exam BOTH System 295038 Exam Submittal Q0 I00 KI/A EK 1.03 RO 2.8 SRO 3.8 Hilgh Off-Site Release Rate IObjective: 29501 LK063 Knowledge of the operational implications of the ]Meteorological effects on off-site release following concepts as they apply to HIGH OFF SITE RELEASE RATE: The following conditions exist at Dresden. A Tornado Warning is in effect for the area that includes Dresden. Reactor Building crane lifts are in progress to move material from the 517 foot level of the reactor building to the refuel floor. Dresden Security personnel have sighted a tornado. Which of the following must be performed as a result of these conditions? A. Start EDG's in anticipation of a loss of off-site power. B. Verify blowout panels are in place on both Unit 2 and 3 Reactor Buildings. C. Open Unit 2 and 3 Turbine Building rollup doors to equalize building pressure. D. Stop crane lifts ONLY if a local assessment determines the tornado will hit on site. Explanation: ANSWER

b. Verify blowout panels are in place on both Unit 2 and 3 Reactor Buildings.

Explanation: Verifying the blowout panels is the required action per DOA 10-2. Answers a and c are items the procedure directs NOT to do and all crane lifts are stopped whether or not the tornado will impact on site. This would affect the off-site release rate due to a loss of secondary containment.

Reference:

Question Pedigree: DOA 00 10-02 Tornado Warning/ Severe Winds New Cog level: Memory Rev 2 03/28/02 I of I

Dresden 2002 1I[" NRC Fxam Q4 Exam System K/A RO SRO 101 RO 300000 K4.02 3.0 3.0 Instrument Air System (IAS) Objective: 278IL-S I-5.d. Knowledge of (INSTRUMENT AIR SYSTEM) Cross-over to other air systems design feature(s) and or interlocks which provide for the following: The Unit 2 Service Air to Unit 2 Instrument Air crosstie will automatically open if. A. service air header pressure decreases to 95 psig B. instrument air header pressure decreases to 85 psig C. radwaste sparging air header pressure decreases to 95 psigy D. control room breathing air header pressure decreases to 85 psig Explanation: ANSWER:

b.

instrument air header pressure decreases to 85 psig Explanation: The crosstie between service air and instrument air opens at 85 psig.

Reference:

SDM278000 Question Pedigree: Bank 27800S0074 Cog level: Memory Rev I 03/28/02 Exam Submittal I of I

Dresden 2002 11 W [\\a'am Q// Exam System A RO SRO 102 SRO 400000 3.33 4 0.4 4.0 Component Cooling Water Systemn (COWS) -7Objective-277 L-S 1-7 During performance of DOS 1500-02 the following pur :- low values were recorded: Unit 2: 2A CCSW pump 3610 gpm Unit 3 3A CCSW pump 3590 gpm B ased u o f Opese in iations ý r c g iize i d c t o sro y t m o e a B.Pa Bot pump re opras becus onl 35e enrylee codtis for: trdfo ah p C. 2A is operable but 3A is NOT because Unit 3 pumps .- st also supply cooling to the Control Roon ventilation. D. 3A is operable but 2A is NOT because Unit 2 pumps -st also supply cooling to the Control Room ventilation. Explanation: ANSWER

d. 3A is operable but 2A is NOT because Unit 2 pumps mu--2 also supply cooling to the Control Room ventilation.

Explanation: Unit 2 pumps must have flow greater than 3621 because th3. also supply cooling to the control room ventilation system.

Reference:

ITS 3.7.1 Bases and DOS 1500-02 (a-estion Pedigree: Cog level: High Rev I 03/28/02 1 of I Exam Submittal clinical

Dresden 2002 II1T NRC Exam Exam BOTH System 500000 K/A EK2.09 High Conta-:-. ent Hydrogen Concentration Knoxvled-e --',ie interrelations between HIGH CONTAIN':-NZNT HYDROGEN CONCENTRA-TIONS the following: Drywell nitrogen purge system During pox e-7 operation, the drywell and torus are normally inerted to... A. allow d-.:ýction of Iodine gas more readily. B. prevent :ie occurence of a flammable mixture in the primary containment. C. control temperatures of the containment during Loss of Coolant Accidents. D. limit the amount of oxygen generated during a LOCA so an explosive mixture is NOT achieved. 03/28/02 Exam Submittal I of I Q# 103 RO 3.0 SRO

3.3 Explanation

ANSWER:

b.

pre% ent the occurrence of a flammable mixture in the primary containment. Explanation: During a LOCA if the drywell and torus were not inerted With nitrogen the build up of hydrogen caused by the zirc-water reaction could lead to a combustible atmosphere in the drywell.

Reference:

223L-S3 Question Pedigree: Modified 22301S0241 Cog level: Memory Rev I I Objiective:223L-S3-1

Dresden 2002) IL" TNRC 'xam System 500000 K/A EK3'.0 I High Containment HIydrogen Concentration Knowledge of the reasons for the following responses as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS: Objective: 29502LK0 IS Initiation of containment atmosphere control system Nitrogen purging of the primary containment without venting while performing the actions of DEOP 200-2 Hydrogen Control will... A. NOT reduce the hydrogen concentration. B. increase the pressure in the containment. 03/28/02 I of I Exam Submittal Q;/ 104 Exam BOTH RO 2.9 SR1O 3.3 C. make the hydrogen monitoring indications unreliable. D. increase the oxygen concentration in the primary containment. Explanation: ANSWER b increase the pressure in the containment. Explanation: Purging the containment without venting will result in pressurizing the containment without lowering the partial pressure or mass of hydrogen.

Reference:

DEOP 500-4 Question Pedigree: New Cog level: Memory Rev 2

Dresden 2002 [L" NRC Exam Exanm RO System 600000 Plant Fire On Site Knowledge of the operation applications of the following concepts as they apply to Plant Fire On Site: 03/28/02 I of Exam Submittal QI 105 K!A A K 1.02 RO 2.9 SRO 3.1

Drcsdcn 2002 II, INRC Fxam System GENERIC K//\\ 2.1.11 Conduct of Operations Objective: 223 L-S4-07 Knowledge of less than one hour technical specification action statements for systems. Unit 3 is operating at rated power. Torus temperature has increased tc -!2)F Operators are required to __ .(I. to ensure that (2) during a DBA LOCA. 2 A. scram the reactor and emerger.;. depressurize the peak primary containment pressures and temperatures do NOT exceed maximum allowable values B. scram the reactor and emergenc-depressurize sufficient net positive suction head is maintained for ECCS pumps C. scram the reactor the peak primary containment pressures and temperatures do NOT exceed maximum allowable values D. scram the reactor sufficient net positive suction head is maintained for ECCS pumps 03/28/02 I of I Exam Submittal Q# 106 Exam SRO RO 3.0 SRO

3.8 Explanation

ANSWER c scram the reactor and the peak pr-nary containment pressures and temperatures do NOT exceed maximum allowable values Explanation: ITS 3.6.2.1 action D requires that N :-h torus average temperature greater than II 0F and reactor power grater than 1%, immediately place the mo:2e switch in shutdown. This is done to prevent the torus from heating up beyond design limits.

Reference:

Question Pedigree: ITS 3.6.2.1 step D and Bases Last years SRO NRC exam #128 (was used for a different KA last year) Cog level: High Rev I Unit 3 is operating at rated power i I

Q[0 107 Exam SRO Dresden 2002 I[LT NRC Exam System K/A GENERIC 2.1.13 RO 2.0 SRO

2.9 Explanation

ANSWER:

c. Nuclear Security Manager Explanation:

Only the Nuclear Security manager can assign an access status level that allows entry into security areas.

Reference:

SY-AA-103-511 Question Pedigree: New Cog level: Memory Rev 2 03/28/02 1 of I Exam S-..rmittal

Dresden 2002 ILT NRC Exam Q# Exam System 1K/A RO SRO 108 BOTH GENERIC 2.1.22 2.8

3.3 Objective

212L-S1-3 Conduct of Operations Ability to determine Mode of Operation. The following conditions exist on Unit 2: The MODE Switch is in the Shutdown position. Reactor coolant temperature is 200'F. Mechanical Maintenance has detensioned two reactor vessel closure bolts. The Reactor is in Mode... A. 2 "Startup" B. 3 "Hot Shutdown" C. 4 "Cold Shutdown" D. 5 "Refueling" Explanation: ANSWER

d. 5 "Refueling" Explanation:

With the reactor shutdown and one or more closure bolts detensioned the reactor is in mode 5.

Reference:

Question Pedigree: ITS Table 1. 1-1I New Cog level: High Rev I 03/28/02 1 of I Exam Submittal

Dresden 2002 I[,T FNRC '- -: Q'/ Ex am System K '.\\ RO SRO 109 RO GENERIC 2.1. 3.4 3.8 ,iective: 21504LK00I Conduct of Operations [Ability to e-. -lain and apply system limits and precautions, DOA 0700-02 SRM OR IRM DETECTOR STUCK, provides. :-ution with regard to the use of portable radios near the SRM/IRM Preamp Cabinet. e What is the reason for this caution? A. Keying the radio may cause a voltage spike resulting in blo.; fuses in the detector circuitry. B. The operation of the radio near the preamnp could cause intek'rence resulting in a reactor scram. C. The operation of the radio near the preamnp could cause circ_:: dampening resulting in lower than actual indication on the SRM/IRM instruments. D. The high voltage that is used by the preamp could cause ex:r.-ne radio interference resulting in poor communications or miscommunication between parties. Explanation: ANSWER

b. The operation of the radio near the preamp could cause interf-ence resulting in a reactor scram.

Explanation: The radio could cause interference and cause a scram, there is a.-te in the surveillance to remind the operators of this.

Reference:

o ues:f)n Pedigree: DOA 700-2 Bank 13000S0011 Cog level: Memory Rev I 03/28/02 1 of I Exam Submittal

Dresden 2002 IIF, NRC Exam 110 3 xam 30FtH System GENERIC K/A 2.1.8 RO 3.8 SRO 3.6 I Objective: 202001,P008 Conduct Of Ope7-7ýo)1s Ability to coordinate personnel activities outside the c oil tro room. U nit 3 is at rate.. - :,nditions. Local adjustme-_: :f Reactor Recirculation pump 3A speed is required. Which of the fo..:,ing describes the MINIMUM requirements to perform this evolution? A. Communici: :i with any on shift Operator prior to adjustment. B. Communica: :n between the Control Room and a licensed Operator at the motor generator. C. Communica: n between the Control Room and on shift Operator at the motor generator with no physical restriction \\ .z:h would prohibit solo operations at the motor generator. D. Communica- :n between the Control Room and an active licensed Operator at the motor generator with no license restrF.:-:on which would prohibit solo operations at the motor generator. 03/28/02 I of [ Exam Submittal Explanation: ANSWER

d. CommunicatiC-between the Control Room and an active licensed Operator at the motor generator with no license restrictio7..vhich would prohibit solo operations at the motor generator..

Explanation: In order to perfor-. Recirc MG Scoop Tube Manual Local Operation, communications must be established between the Con7-.1 Room and the operator at the applicable recirc MG set. The operator at the MG set must have an active lic -se with no license restrictions.

Reference:

Question Pedigree: DOP 0202-12 Modified last years RO NRC Exam #110 Cog level: Memo-, Rev 2

Dresden 2002 ILT NRC Exam System GENERIC K/A 2.2.2 RO

4.0 Objective

215L002-02 Ability to manipulate the console controls as required operate the facility between shutdown and designated power levels. I of I Exam Submittal Q# Exam BOTH SRO 3.5 03/28/02

[)resden 2002 1lT NRC Exam 112 Exam BO tlI System GENERIC 2.2.26 RO 2.5 A. The Control Room Nuclear Observer is in the Control Room. B. Radiation Protection Personnel have placed a high radiation area lock on AND posted the acce s ladders to the Drywell above the first floor indicating: NO ENTRY FUEL TRANSFER IN PROGRESS C. A SRO or an SROL is directly supervising and in line of sight of fuel handling operations on Refueling Platform. D. A Qualified Nuclear Engineer verifies SRM reading are as expected after each step of the Nu,'ear Component Transfer List. CO Explanation: ANSWER

c. A SRO or an SROL is directly supervising and in line of sight of fuel handling operations on th. Refueling Platform.

Explanation: DFP requires that an SRO or SROL be in line of sight of fuel handling operation on the refuel floe.- during core alterations.

Reference:

Question Pedigree: DFP 800-1 and TRM 3.9.a New Cog level: Memory Rev 2 03/28/02 I of I Exam Submittal 57 ) 2 - O)jcctive: 29800LI,,084 K 014 Equipment Control Of refueling administrative requi.-- 7ionts. ""cct' 'e cot n- ý be Iniet? core alterations that potentially affect core reactivity which off t1hee ffollowing conditions 7n be met?

Dresden. .-. LT NRC Exalml Q/1 113 03/28/02 Exam Submittal Exam SRO System GENERIC K/A 2.2.3 RO 3.1 SRO 3.3 I c."

Dresden 2002 [L.TI NRC Exam System GENERIC K/A 2.2.34 Equi-mnent Control Objective: 20102LK032 Knowledge of the process for determining the internal and external effects on core reactivity. Unit ' is near the end of an operating cycle with a startup in progress. Reac-cor coolant temperature has lowered 30°F below the value that was used to calculate the ECP. Who :s required to recalculate the ECP? A. Nuclear Station Operator B. Unit Supervisor C. Shift Technical Advisor D. Qualified Nuclear Engineer Explanation: ANSWER d Qualified Nuclear Engineer Explanation: At the end of an operating cycle the potential exists for the moderator temperature coefficient to be positive at low moderator temperatures. The QNEs assistance is required to make this determination

Reference:

DGP 1-1 Question Pedigree: New Cog level: Memory Rev I 03!28,'02-1 of I Exam Submi Q11 114 Emall BOTf I RO 2.8 S RO 3.2 tt a

Dresden 2002 ILT NRC Exam Exam SRO System GENERIC Exam Submittal 115 K/A 2.2.8 RO0 1.8 SRO

3.3 FObjfective

29900,K 108 qumpment Uontroi Knowledge of the process for determining if the proposed Knowledge of the process for determining if the proposed change, test, or experiment involves an unreviewed safety question. A systems engineer brings a "Special Procedure" for the Control Rod Drive system to the WEC. How can it be determined if the "Special Procedure" contains any unreviewed safety questions? A. The signature on the procedure by the system engineer approving the special procedure. B. The Shift Manager informed the crew at turnover the special procedure was scheduled to be performed. C. The special procedure has been screened in accordance with OP-AA-101-304 "Evaluation of Special Tests or Evolutions". D. The documentation of a 50.59 review being conducted on the special procedure is included with the special procedure. Explanation: ANSWER

d. The documentation of a 50.59 review being conducted on the special procedure is included with the special procedure.

Explanation: SRO only criteria #3 10 CFR 50.59 requires a written Safety Evaluation which provides the bases for the determination that the change, test, or experiment does not involve an Unreviewed Safety Question.

Reference:

LS-AA-999 Section A Purpose Question Pedigree: New Cog level: Memory Rev I 03/28/02 I of I inquipmnent Control

Dresden 2002 ILF NRC Exam Q# Exam System K/A RO SRO 116 RO GENERIC 2.3.1 26

3.0 Objective

Radiological Controls Knowledge of 10 CFR 20 ar -ated facility radiation control requirements. The RWCU pump room was recently surveyed and the following radiological co- _-.ions exist: General area radiation of 20 mRem per hour Smearable contamination of 100 dpm/100 cm2 (beta-gamma) Which of the following postings should be applied to this area? A. Radiation area only B. High radiation area only C. Radiation area and Contamination area D. High radiation area and Contamination area Explanation: ANSWER

a. Radiation area only Explanation:

A Radiation Area is any area within an RPA accessible to individuals, in which rac -:ion levels from radiation sources external to the body could result in an individual receiving a deep dose eqe --, alent greater than 5 mrem/hr but less than 100 mrem/hr at 30 centimeters from the radiation source or ::- rn any surface that radiation penetrates. A Contamination Area has smearable contamination greater t-10OOdpm/100 cm 2.

Reference:

RP-AA-376 Question Pedigree: Modified 29400S0261 Modified Last years NR 7 exam # 116 Cog level: High Rev 2 03/28/02 1 of I Exam Submittal

Dresden 2002 II [ NRC lxam Q# Exam System KiA RO SRO 117 RO GENERIC 2.3.10 2.9 3.3 R Objective: 29501 LK051 Radiological Controls Ability to perform procedures to reduce excessive levels of radiation and guard against personnel exposure. Chemistry has reported that higa -.:olant activity exists on Unit 2 and a fuel element failure is suspected. Which of the following actions ;;z- :quired to prevent excessive personnel exposure if site assembly is required? A. Isolating HPCI steam drains B. Isolating the Isolation Conde-,er C. Isolating HPCI steam flow D. Isolating Recirc Sample Linesý Explanation: ANSWER

a. Isolating HPCI steam drains Explanation:

Assembly area inside the RPA is 7-ear the feedpumps, which is against the condenser shield wall. Any flow of radioactive water to the condenser wxould increase dose rates in this area, so a is correct. IC drains do not go to the condenser, so b is incorrect. Isl*ating HPCI steam flow would block the leakage, but would render HPCI unavailable, so this action would r. Dt be appropriate. Recirc sample drains also do not go to the condenser.

Reference:

DGA-16 section D. I I Caution. Question Pedigree: 1998 NRC Exam Cog level: Memory Rev I 03/28/02 1 of I Exam Submittal

Dresden 2002 II,T NRC Fxam System GENERIC K/A I of I Exam Submittal Q# 118 Ex-aII BOTI I RO 2.5 SRO 2.9 03/28/02

Dresden 2002 ILT NIRC Exam Q# Exam System K/A RO S RO 119 BOTI-[ GENERIC 2.3.4 2.5

3.1 SObjective

NGETF Radiological Controls Knwegeo adiation exposure limits and contaminato Control, including, permissible levels in excess of those au th orized, Per RP-AA-203, Exposure Review and Authorization, workers at Dresden have a administrative exposure control level of (1) mrem TEDE per year. This can be raised to (2) mrem TEDE by the Radiation Protection Manager. 1 2 A. 1000 3000 B. 1000 5000 C. 2000 3000 D. 2000 5000 Explanation: ANSWER

c. 2000 and 3000 Explanation:

Workers at Dresden have a limit of 2000 mrem per year the RP manager can raise this to 3000.

Reference:

Question Pedigree: RP-AA-203 New Cog level: High Rev 2 03/28/02 Exam Submittal I of]

Dresden 2002 ILT VNRc: Q# Exam System K A RO SRO 120 SRO GENERIC 2.,.(, 2.1 3.1 -lective-272L-S2-0 Radiological Controls Knowledge :W .ie requirements for reviewing an( approving - -.se pern its. DOP 2000-110, Attachment 1: Waste Surge Tank Radioactive - charge to River Card, contains tile calculation for determining the f1) flowrate and ra agical monitor alarm setpoints, and verified by. ffi 1 2 A. discharge Field S _:ervisor B. discharge Shift N.a-.ager or designee C. dilution Field S--ervisor D. dilution Shift M -Žager or designee Explanation: ANSWER B discharge and SM or designee Explanation: The discharge flow rate is determine and verified by the SM or dc-:nee.

Reference:

268N-03, DOP 2000-110 Queso edigree: Last

2:-s NRC exam #119 Cog level: Memory Rev I 03/28/02 I of I Exam Submittal

Dresden 2002 ILT NRC Exam QExam System K/A RO SRO 121 SRO GENERIC 2.3.9 2.5

3.4 Objective

29502LK068 Radiological C -:-ols Knowledge of the process for performing a containment purge. The drywell is,-ig vented to control H 2 and 02. The following -:-s were noted prior to initiating venting: Torus level is `- et. Drywell Torus Hydreo.2 7% 7% Oxyge-7 7% 7% After some peri:,- of time, it is determined that drywell hydrogen and oxygen cannot be controlled with SBGT and Nitrogen PL-ý;. In this condition.ýhich of the following is the proper response? A. Immediatel-, zpray the torus B. Begin simu::.--n.eous venting of the torus AND drywell. C. Vent and pL7-ge the containment per DEOP 500-4 "Containment Venting", Attachment 4. D. Vent and pu:_e the containment per DEOP 500-4 "Containment Venting", Attachment 5. Explanation: ANSWER c Vent and purge -:e containment per DEOP 500-4 "Containment Venting", Attachment 4. Explanation Only Attachmen: I addresses venting the drywell through the RB ventilation system. With the concentrations given torus spray are not allowed. Simultaneous venting of the torus and drywell is never done. Torus level is too high to allox;, enting of the Torus. Provide DEOP 0-2 and DEOP 500-4 with Section H blanked out.

Reference:

DEO-200-2 and 500-4 Question Pedigree: Modified last years SRO only NRC exam #127 Cog level: High Rev I 03/28/02 1 of I Exam Submittal

[)resden 2002 [ILT NRC Exam QI/ Exam System K/A RO SRO 122 SRO GENERIC 2.4.32 3.3

3.5 Objective

29501LP059 Emergency Procedures and Plan Knowledge of operator response to loss of all annunciators. Unit 2 was at rated conditions when "ANNUN DC PWR FAILURE" alarms are received on several panels simultaneously. A bell inside 902-4 sounds Which of the following describes the expected operator actions? A. Scram the reactor due to the loss of annunciators. The Shift Manager should evaluate for a possible GSEP condition. B. Verify that the nonnal AC power supply is still available by performing an annunciator checks on each effected panel. Notification of the Shift Manager is NOT required. C. Verify that the normal AC power supply is still available by performing an annunciator checks on each effected panel. Notification of the Shift Manager is required. D. Determine the cause of the loss of annunciators. The Shift Manager should evaluate for a possible GSEP condition. [Explanatioon: ANSWER

e.

Determine the cause of the loss of annunciators. The Shift Manager should evaluate for a possible GSEP condition. Explanation: Receipt of these alarms indicates a failure of the panels Annunciator System. Operators should determine the cause of the loss of annunciators and attempt to restore. The Shift Manager should evaluate for a possible GSEP condition.

Reference:

DAN 902(3)-5 H-3 Question Pedigree: Last years NRC exam #122 Cog level: High Rev 2 03/28/02 1 of I Exam Submittal

E xan) BOTH Dresden 2002 ILT NRC Exam Y GE tcLI) K/A RO( ,'0 ENERIC 2.4.35 Objective: 20101Lp010 Knowledge of local auxiliary operator task, Lurin-g emergency operations including system gea :aphy and system implications. control rod drive, the operator wii! need a hose. adid<- ..r Explanation: ANSWER:

d.

Turbine building second floor in the DEOP storage locker Explanation: The tools and equipment needed to vent the overpiston area of the CRDs is located in the DEOP s:)rage locker in the TB between the Units near the TBCCW HX.

Reference:

DEOP 500-05 Question Pedigree: Bank 29502S0521 Cog level: Memory Rev I 03/28/02 Exam Submittal I of I Q/1 123

Q# 124 Exam BOTH 03/28/02 Exam Submittal Dresden-HA' NRC Fxam System K/A GENER' 2.4.45 i : 1 RO 3.3 SRO

3.6 Objective

29900LK042 Emergency Procedures and Plan Ability to prioritize and interpret the significance of each annunciator or alarm. During normal plant operations, an annunciator -rms which has a RED backlight. What is the significance of this RED backlighr A. Identifies a parameter which could cause a _- : scram. B. Identifies a parameter which causes a critic. :.iange in plant status. C. Identifies a condition that requires immedia-: entry into the DOA's or DEOP's. D. Informs the operators of annunciators that ax- ýxpected to alarm due to maintenance being performed. Explanation: ANSWER:

b.

Identifies a parameter which causes a cr:: :al change in plant status. Explanation: The backlit red annunciators do not all identify a -: ndition that could cause a scram or entry into a DOA or DEOP. They also do not indicate when maintena- :r is being performed.

Reference:

Operator Aid #84 Question Pedigree: Bank 29902S0401 Cog level: Memory Rev I

Dresden 2002 ILT NRC Exam Exam BOTII System GENERIC Exam Submit>:a K/A 2.4.7 RO 3.! SRO

3.8 SObjective

29502LK003) Ennergency Procedures and Plan Knowledge of event based EOP mitigation strategies. DEOP 200-01 requires emergency depressurization if torus water level cannot be maintained above 11 ft. \\.nat is the reason for this action? A. T-Quenchers are uncovered at 10.8 ft B. The loss of HPCI requires Low Pressure ECCS injection C. Reject energy from the vessel while the suppression pool is still available D. Torus water volume is too low to absorb RPV energy with pressure greater than 1100 psig Explanation: A\\NSWER: C reject energy from vessel while suppression pool still available Explanation: Emergency Depressurize must be performed if level cannot be maintained above I Ift since that's the level of the downcomers. If a LOCA occurred, with level below II ft, the steam would not be discharged under water and would pressurize the Torus Air Space.

Reference:

295L-S2 Question Pedigree: Bank 29502S0271 Cog level: Memory Rev 1 03!'28 02 I of I}}